JPH038714B2 - - Google Patents
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- Publication number
- JPH038714B2 JPH038714B2 JP59013063A JP1306384A JPH038714B2 JP H038714 B2 JPH038714 B2 JP H038714B2 JP 59013063 A JP59013063 A JP 59013063A JP 1306384 A JP1306384 A JP 1306384A JP H038714 B2 JPH038714 B2 JP H038714B2
- Authority
- JP
- Japan
- Prior art keywords
- tritium
- plasma
- wall
- structural material
- core structural
- 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 - Lifetime
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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
- Y02E30/10—Nuclear fusion reactors
Landscapes
- Plasma Technology (AREA)
- Particle Accelerators (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Description
本発明は核融合炉の炉心構造材へのトリチウム
溶解量及びトリチウム透過量の低減方法に関す
る。
核融合炉、例えば円周方向に沿つて作つたプラ
ズマ中に電流を流して、この電流によつて発生す
るポロイダル磁場の力によつてプラズマをトーラ
ス状に閉じこめて核融合反応を起させるトカマク
型核融合炉では、第１図に示すごとくドーナツ状
のプラズマ１を囲むように内側ブランケツト２、
外側ブランケツト２′、内側遮蔽体３、外側遮蔽
体３′、プラズマを磁力線の力で安定して閉込め
るため及び磁力線に端のない円周方向の磁場を作
るためのトロイダル磁場コイル４、ボロイダル磁
場コイル５等が配置されている。その他高いエネ
ルギーを持つた電気的に中性の粒子をプラズマに
入射して、そのエネルギーをプラズマに与えるこ
とによつて、プラズマの温度を上げる中性粒子入
射加熱装置（図示省略）、プラズマの周辺の磁力
線の形状に工夫して、外に逃げ出したプラズマが
直接内側ブランケツト２の壁に当らないように
し、特にプラズマの中の不純物を減少させるのに
効果があるダイバータ６、プラズマ中に混入して
くる不純物や核融合反応の生成物であるヘリウム
を系外に排出する排気装置７等が設置されてい
る。トカマク型核融合炉は概ね上記のように構成
されているが、このうち内側及び外側ブランケツ
ト２，２′は重要な構成機器であり、容器内のリ
チウム化合物と融合反応により発生した中性子と
を反応せしめて、核融合炉の燃料となるトリチウ
ム（三重水素）を生産する機能、その中性子のも
つエネルギーを熱エネルギーに変換する機能、更
には内側及び外側遮弊体３，３′と共に中性子の
放射線遮蔽をする機能を備えている。プラズマ１
を取り囲む内側、外側ブラケケツト２，２′のプ
ラズマ１に対向する壁である第１壁及びダイバー
タ６等の炉心構成材は、プラズマ領域から漏洩し
て来る高速中性子、荷電／中性トリチウム粒子及
びヘリウム粒子等の厳しい照射を受ける。特に荷
電／中性トリチウム粒子は高温の構造材中では高
透過性を有し核材料中に大量のトリチウムが溶解
することから、冷却水に拡散透過する。
従来、第１壁などの炉心構造材は第２図に示す
ように冷却パネルで構成されていて、プラズマ対
向壁８の裏面側に冷却材流路９を設け、該流路９
の中にH2OあるいはD2O等の冷却材１０を通し
て、炉心プラズマから放射される高速中性子束、
荷電／中性トリチウム束等の熱負荷及び粒子負荷
によつて過熱されるプラズマ対向壁８の材料温度
を許容温度以下に保持していた。この時プラズマ
対向壁８の表面より入射した荷電／中性トリチウ
ム粒子の大半は拡散と表面での再結合反応（Ｔ＋
Ｔ→T2）を経てT2分子の形態で再放出されるが、
その入射トリチウムの輸送機構の概念は第６図に
示す通りである。この再放出過程において、拡散
距離が非常に短いため（数＋Å）、表面再結合が
律速過程となるので、表面近傍でのトリチウム最
大濃度Ｃはプラズマ１より入射して来るトリチウ
ムの入射粒子束Ｊと表面再結合の反応率のバラン
スで決り、その関係は(1)式のとおりである。
ＪKC2 ………(1)
ここでＫは再結合定数である。一方表面近傍の
最大濃度Ｃに起因する濃度勾配によつて、トリチ
ウムは濃度の小さい冷却水側に拡散透過する。こ
のように大量のトリチウムが炉心構造材中に溶解
し、冷却水にまで拡散透過することは、トリチウ
ムが放射性核種であることから、分解修理時の作
業環境と冷却水汚染の安全性に問題があつた。さ
らにトリチウムが大量に溶解して、トリチウムイ
ンベントリーが増大するため、運転初期に必要な
燃料トリチウム量の確保に影響を与えていた。
本発明は、斯かる問題を解消すべくなされたも
のであり、再結合定数の大きな材料のコーテイン
グによつて核融合炉の炉心構造材へのトリチウム
溶解量及びトリチウム透過量の低減方法を提供せ
んとするものである。
本発明のトリチウム溶解量及びトリチウム透過
量の低減方法は、核融合炉のプラズマに対向する
壁の裏面側に、該プラズマ対向壁の構成材料の温
度が許容温度以下になるよう冷却するための冷却
材を通す冷却材流路管を設けてなるブランケツト
の第１壁及びダイバータの炉心構造材に於て、該
構造材のプラズマ側表面に、水素の表面再結合反
応（Ｈ＋Ｈ→H2）の速度定数（式(1)の再結合定
数Ｋ）の大きい金属をコーテイングして、該構造
材の中に入射してくるトリチウムの表面での再放
出を促進することを特徴とするものである。
以下本発明の一実施例について詳細に説明す
る。第３図は支持構造を含んだブランケツトの構
造概念の斜視図で代表例として外側ブランケツト
２′を示している。８は第１壁であるプラズマ対
向壁、１１は遮蔽体、１２，１３はブランケツ
ト、遮蔽体の支持部材である。プラズマ対向壁８
の詳細構造は第４図に示す通りで、プラズマ対向
壁８の表面にAl、フエライト鋼等のトリチウム
再結合定数の大きな金属１４をコーテイングして
あり、背面側にコーテイング材１４及び該プラズ
マ対向壁８の構造材がプラズマから厳しい熱負荷
と粒子負荷を受けて過熱するのを防止するための
冷却材１０を通す冷却材流路９が設けられてい
る。一方、もう一つの炉心構造材であるダイバー
タの構造概念は第５図の斜視図に示した通りであ
るが、内側ブランケツト２の排気流路の入口側に
おいてプラズマ対向壁８の表面側に設置される。
１５はダイバータ板、１６は該ダイバータ板を冷
却するための冷却配管で、１７は入口ヘツダー、
１８は出口ヘツダーであり、該出入口ヘツダー１
７，１８に接続される出入口配管は図示を省略し
たが遮蔽体を貫通して炉外に導びかれている。
次に上記の如く構成した本実施例の作用を説明
する。関係式(1)から見られるように表面再結合定
数Ｋを大きくすれば、表面近傍のトリチウム濃度
Ｃは小くなり溶解量及び透過量を低減することが
可能となる。第７図に各種材料の再結合定数（再
結合反応速度定数）Ｋの理論値を示す。縦軸にＫ
値、横軸に温度（〓）の逆数を1000倍にして整理
した値である。第１壁及びダイバータ板の構造材
は通常工業的信頼性の高いオーステナイト・ステ
ンレス鋼が使用されているが、第７図に見られる
ようにＫ値はAl、フエライト鋼などよりかなり
小さい材料である。しかしながらＫ値の大きい材
料は材料強度が小さく、高温強度材料としては適
していないので、本発明では第１壁及びダイバー
タの炉心構造材は高温強度の強いオーステナイ
ト・ステンレス鋼等の材料で構成し、そのプラズ
マ側表面にＫ値の大きい材料である、Al、Cu等
をコーテイングするかあるいは該材料の薄板を圧
着、爆着等によつてはり合せた材料で構成してい
る。次表に第１壁条件（例えば180℃）での
SUS、Al、Cu、Fe（bcc）の表面での再結合定数
Ｋ、SUSのＫ値との比Ｋ／KSUS、及び表面近傍で
のトリチウム濃度ＣとSUSのＣ値との比Ｃ／CSUS
を示す。
The present invention relates to a method for reducing the amount of tritium dissolved and the amount of tritium permeated into core structural materials of a nuclear fusion reactor. A nuclear fusion reactor, for example, a tokamak type, in which a current is passed through the plasma created along the circumference, and the force of the poloidal magnetic field generated by this current confines the plasma in a torus shape to cause a nuclear fusion reaction. In a fusion reactor, as shown in Figure 1, an inner blanket 2,
Outer blanket 2', inner shield 3, outer shield 3', toroidal magnetic field coil 4 for stably confining plasma with the force of magnetic lines of force and for creating a circumferential magnetic field with no ends in magnetic lines of force, voloidal magnetic field. A coil 5 etc. are arranged. A neutral particle injection heating device (not shown) that increases the temperature of the plasma by injecting electrically neutral particles with high energy into the plasma and imparting that energy to the plasma; The shape of the magnetic field lines is designed to prevent the plasma escaping to the outside from directly hitting the wall of the inner blanket 2, and the diverter 6, which is particularly effective in reducing impurities in the plasma, prevents the plasma from entering the plasma. An exhaust device 7 and the like are installed to exhaust helium, which is a product of nuclear fusion reactions and impurities generated by nuclear fusion reactions, out of the system. A tokamak-type fusion reactor is generally configured as described above, but the inner and outer blankets 2 and 2' are important components, and are used to react between the lithium compound in the container and the neutrons generated by the fusion reaction. At the very least, there is a function to produce tritium (tritium), which is the fuel for a nuclear fusion reactor, a function to convert the energy of neutrons into thermal energy, and a function to shield neutron radiation with the inner and outer shields 3 and 3'. It has the ability to plasma 1
The first wall, which is the wall facing the plasma 1 of the inner and outer brackets 2 and 2' surrounding the reactor core components such as the diverter 6, absorbs fast neutrons, charged/neutral tritium particles, and helium leaking from the plasma region. Subject to severe irradiation from particles, etc. In particular, charged/neutral tritium particles have high permeability in high-temperature structural materials, and a large amount of tritium dissolves in the core material, so that they diffuse into the cooling water. Conventionally, core structural members such as the first wall are composed of cooling panels as shown in FIG.
Fast neutron flux emitted from the core plasma through a coolant 10 such as H 2 O or D 2 O,
The temperature of the material of the plasma facing wall 8, which is overheated by the heat load and particle load of charged/neutral tritium bundles, was kept below the allowable temperature. At this time, most of the charged/neutral tritium particles incident from the surface of the plasma facing wall 8 undergo diffusion and surface recombination reaction (T+
T → T 2 ) and is re-released in the form of T 2 molecules,
The concept of the transport mechanism of incident tritium is as shown in FIG. In this re-emission process, since the diffusion distance is very short (several + Å), surface recombination becomes the rate-determining process. It is determined by the balance between the surface recombination reaction rate and the surface recombination reaction rate, and the relationship is as shown in equation (1). JKC 2 ......(1) Here, K is the recombination constant. On the other hand, due to the concentration gradient caused by the maximum concentration C near the surface, tritium diffuses and permeates toward the cooling water side where the concentration is lower. Since tritium is a radionuclide, dissolving a large amount of tritium in the core structural materials and diffusing into the cooling water poses problems to the working environment during overhaul and to the safety of cooling water contamination. It was hot. Furthermore, a large amount of tritium was dissolved, increasing the tritium inventory, which affected the ability to secure the amount of fuel tritium needed in the early stages of operation. The present invention has been made to solve this problem, and provides a method for reducing the amount of tritium dissolved and the amount of tritium permeated into the core structural material of a fusion reactor by coating with a material having a large recombination constant. That is. The method for reducing the amount of tritium dissolved and the amount of tritium permeated according to the present invention is to provide cooling to the back side of the wall facing the plasma of the fusion reactor so that the temperature of the constituent material of the plasma facing wall becomes below the permissible temperature. In the first wall of the blanket provided with the coolant flow pipe through which the material passes, and in the core structural material of the divertor, the surface recombination reaction of hydrogen (H+H→H 2 ) is observed on the plasma side surface of the structural material. It is characterized in that it is coated with a metal having a large constant (recombination constant K in formula (1)) to promote re-emission of tritium incident on the surface of the structural material. An embodiment of the present invention will be described in detail below. FIG. 3 is a perspective view of the structural concept of the blanket including the support structure, and shows the outer blanket 2' as a representative example. Reference numeral 8 denotes a plasma facing wall which is a first wall, 11 a shielding body, 12 and 13 a blanket and supporting members for the shielding body. Plasma facing wall 8
The detailed structure is as shown in FIG. 4, in which the surface of the plasma facing wall 8 is coated with a metal 14 having a large tritium recombination constant such as Al or ferrite steel, and the coating material 14 and the plasma facing wall are coated on the back side. A coolant flow path 9 is provided through which a coolant 10 passes to prevent the structural members 8 from overheating due to severe heat and particle loads from the plasma. On the other hand, the structural concept of the diverter, which is another core structural member, is as shown in the perspective view of FIG. Ru.
15 is a diverter plate, 16 is a cooling pipe for cooling the diverter plate, 17 is an inlet header,
18 is an exit header, and the entrance header 1
Although not shown, the inlet/outlet pipes connected to 7 and 18 are led out of the furnace through the shield. Next, the operation of this embodiment configured as described above will be explained. As can be seen from the relational expression (1), if the surface recombination constant K is increased, the tritium concentration C near the surface becomes smaller, making it possible to reduce the amount of dissolution and the amount of permeation. FIG. 7 shows the theoretical values of the recombination constant (recombination reaction rate constant) K of various materials. K on the vertical axis
The values are arranged by multiplying the reciprocal of temperature (〓) by 1000 on the horizontal axis. The structural material of the first wall and diverter plate is usually austenitic stainless steel, which has high industrial reliability, but as shown in Figure 7, the K value of this material is much smaller than that of Al, ferrite steel, etc. . However, materials with a large K value have low material strength and are not suitable as high-temperature strength materials. Therefore, in the present invention, the core structural materials of the first wall and divertor are made of materials such as austenitic stainless steel that have strong high-temperature strength. The surface on the plasma side is coated with a material having a large K value, such as Al or Cu, or is made of a material in which thin plates of the material are bonded together by pressure bonding, explosion bonding, or the like. The table below shows the conditions for the first wall (e.g. 180℃).
Recombination constant K at the surface of SUS, Al, Cu, Fe (bcc), the ratio K/K to the K value of SUS , and the ratio C/C between the tritium concentration C near the surface and the C value of SUS SUS
shows.
【表】
このようにAl、Cu等の再結合定数の大きな材
料をコーテイングすることによつて、表面再結合
反応（Ｔ＋Ｔ→T2）を促進し、表面トリチウム
濃度Ｃを低下させることが可能となつた。
以上詳述した通り本発明のトリチウム溶解量及
びトリチウム透過量の低減方法は、第１壁、ダイ
バータの炉心構造材のプラズマ側表面にトリチウ
ムの表面再結合定数の大きな材料をコーテイング
するのであるから、該炉心構造材に入射して来る
トリチウムの炉心構造材表面での再放出が促進さ
れ、炉心構造材のトリチウム濃度を低減すること
が可能で、従つて冷却材中へのトリチウムの拡散
透過を抑止することができるという優れた効果を
奏する。[Table] By coating materials with large recombination constants such as Al and Cu, it is possible to promote the surface recombination reaction (T+T→T 2 ) and reduce the surface tritium concentration C. Summer. As detailed above, the method for reducing the amount of tritium dissolved and the amount of tritium transmitted according to the present invention coats the first wall and the plasma side surface of the divertor core structural material with a material having a large tritium surface recombination constant. The re-release of tritium that has entered the core structural material on the surface of the core structural material is promoted, making it possible to reduce the tritium concentration in the core structural material, and thus inhibiting the diffusion and permeation of tritium into the coolant. It has the excellent effect of being able to
第１図はトカマク型核融合炉の概略を示す断面
図、第２図は従来のブランケツト第１壁の構造を
示す斜視図、第３図はブランケツトの概念を示す
斜視図、第４図は本発明の一実施例を示すブラン
ケツト第１壁の構造を示す斜視図、第５図はダイ
バータの概念を示す斜視図、第６図は入射トリチ
ウムの輸送機構の概念図、第７図は再結合反応速
度定数を示すグラフである。
８……プラズマ対向壁、９……冷却材流路、１
０……冷却材、１１……遮蔽体、１２，１３……
支持部材、１４……コーテイング材。
Figure 1 is a cross-sectional view showing the outline of a tokamak-type fusion reactor, Figure 2 is a perspective view showing the structure of the first wall of a conventional blanket, Figure 3 is a perspective view showing the concept of the blanket, and Figure 4 is a diagram showing the structure of the first wall of a conventional blanket. A perspective view showing the structure of the first wall of the blanket showing one embodiment of the invention, FIG. 5 a perspective view showing the concept of a diverter, FIG. 6 a conceptual diagram of the transport mechanism of incident tritium, and FIG. 7 showing the recombination reaction. 1 is a graph showing rate constants. 8... Plasma facing wall, 9... Coolant channel, 1
0... Coolant, 11... Shielding body, 12, 13...
Supporting member, 14... coating material.
Claims (1)
を設けてなるブランケツトの第１壁及びダイバー
タの炉心構造材において、該炉心構造材のプラズ
マ側表面に水素の表面再結合反応の速度定数の大
きい金属をコーテイングして、該炉心構造材中に
入射して来るトリチウムの炉心構造材表面での再
放出を促進することを特徴とする核融合炉の炉心
構造材へのトリチウム溶解量及びトリチウム透過
量の低減方法。1. In the first wall of the blanket, which has a coolant flow path on the back side of the wall facing the plasma, and in the core structural material of the diverter, the rate constant of the surface recombination reaction of hydrogen is applied to the plasma side surface of the core structural material. Tritium dissolution amount and tritium permeation into a core structural material of a nuclear fusion reactor characterized by coating with a large metal to promote re-emission of tritium incident into the core structural material on the surface of the core structural material. How to reduce the amount.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP59013063A JPS60157073A (en) | 1984-01-27 | 1984-01-27 | Method of reducing quantity of tritium melted and quantity of tritium permeated to core structure material of fusion reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP59013063A JPS60157073A (en) | 1984-01-27 | 1984-01-27 | Method of reducing quantity of tritium melted and quantity of tritium permeated to core structure material of fusion reactor |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS60157073A JPS60157073A (en) | 1985-08-17 |
| JPH038714B2 true JPH038714B2 (en) | 1991-02-06 |
Family
ID=11822677
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP59013063A Granted JPS60157073A (en) | 1984-01-27 | 1984-01-27 | Method of reducing quantity of tritium melted and quantity of tritium permeated to core structure material of fusion reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS60157073A (en) |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB202020283D0 (en) | 2020-12-21 | 2021-02-03 | Tokamak Energy Ltd | Divertor cooling |
Family Cites Families (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS56115983A (en) * | 1980-02-19 | 1981-09-11 | Tokyo Shibaura Electric Co | Nuclear fusion reactor |
| JPS57196187A (en) * | 1981-05-29 | 1982-12-02 | Mitsubishi Atomic Power Ind | Plasma incident plate |
-
1984
- 1984-01-27 JP JP59013063A patent/JPS60157073A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS60157073A (en) | 1985-08-17 |
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