JPS6364754B2 - - Google Patents
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- Publication number
- JPS6364754B2 JPS6364754B2 JP56074897A JP7489781A JPS6364754B2 JP S6364754 B2 JPS6364754 B2 JP S6364754B2 JP 56074897 A JP56074897 A JP 56074897A JP 7489781 A JP7489781 A JP 7489781A JP S6364754 B2 JPS6364754 B2 JP S6364754B2
- Authority
- JP
- Japan
- Prior art keywords
- molten salt
- fissile material
- nuclear
- salt
- paragraph
- 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
- 150000003839 salts Chemical class 0.000 claims description 51
- 239000000463 material Substances 0.000 claims description 27
- 238000006243 chemical reaction Methods 0.000 claims description 17
- 238000000034 method Methods 0.000 claims description 11
- PXHVJJICTQNCMI-UHFFFAOYSA-N Nickel Chemical compound [Ni] PXHVJJICTQNCMI-UHFFFAOYSA-N 0.000 claims description 8
- 229910001338 liquidmetal Inorganic materials 0.000 claims description 4
- 229910052759 nickel Inorganic materials 0.000 claims description 4
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical compound [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 claims description 3
- 239000012530 fluid Substances 0.000 claims description 3
- 229910002804 graphite Inorganic materials 0.000 claims description 3
- 239000010439 graphite Substances 0.000 claims description 3
- 229910000856 hastalloy Inorganic materials 0.000 claims 1
- 239000007788 liquid Substances 0.000 claims 1
- 239000002245 particle Substances 0.000 claims 1
- 239000000126 substance Substances 0.000 claims 1
- 230000004992 fission Effects 0.000 description 5
- 238000004519 manufacturing process Methods 0.000 description 5
- 239000000446 fuel Substances 0.000 description 3
- 239000012535 impurity Substances 0.000 description 3
- 210000002268 wool Anatomy 0.000 description 3
- 238000000746 purification Methods 0.000 description 2
- 229910004366 ThF4 Inorganic materials 0.000 description 1
- 239000000654 additive Substances 0.000 description 1
- 230000000996 additive effect Effects 0.000 description 1
- 230000007423 decrease Effects 0.000 description 1
- 238000010586 diagram Methods 0.000 description 1
- 230000020169 heat generation Effects 0.000 description 1
- 239000003758 nuclear fuel Substances 0.000 description 1
- 238000010248 power generation Methods 0.000 description 1
- 238000007670 refining Methods 0.000 description 1
- 239000007787 solid Substances 0.000 description 1
- 239000002699 waste material Substances 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
- Manufacture And Refinement Of Metals (AREA)
Description
【発明の詳細な説明】
本発明は溶融塩核化学反応方法及び装置に関す
る。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a molten salt nuclear chemical reaction method and apparatus.
本願発明者等は、さきに単一流体型加速器溶融
塩増殖炉（以下、sf AMSBと略称する）を発明
し、特許出願した（特願昭55−123552号（特開昭
57−48685号））。このsf AMSBは核分裂性物質
なしに、核分裂性物質を生産することを目的と
し、核分裂性物質がなくてもTh（またはＵ）のみ
があれば 233U（または 239Pu）を生産できる方法
などを提供したが、生産された 233Uの濃度は１
年間の炉運転後でも0.1m／ｏ 233UF4以下であ
り、多量の核分裂性物質インベントリーを持つこ
とは考えなかつた。しかしながら、 233Uが次第
に蓄積されてくると、これを利用してさらに高利
得で経済的なAMSBを設計運転できる可能性が
ある。すなはち、残留された核分裂性物質はそれ
自身の核分裂によつてそれに使はれた中性子が約
2.5倍に増倍され、それにより、全体として、核
分裂性物質の生産効率が高められる。また、核分
裂が増えることによつて、発熱量、従つて発電量
の増大があり、それだけ自家消費電力の確保及び
余剰電力の提供がより確実となる。 The inventors of this application previously invented a single fluid accelerator molten salt breeder reactor (hereinafter abbreviated as sf AMSB) and filed a patent application (Japanese Patent Application No.
No. 57-48685)). The purpose of this SF AMSB is to produce fissile material without fissile material, and we have developed a method that can produce 233 U (or 239 Pu) with only Th (or U) without fissile material. However, the concentration of 233 U produced was 1
Even after one year of reactor operation, it was less than 0.1 m/o 233 UF 4 , and we did not consider having a large inventory of fissile material. However, as 233 U gradually accumulates, it may be possible to use it to design and operate even higher-gain, more economical AMSBs. In other words, the remaining fissile material loses the neutrons used in it by its own fission.
It is multiplied by 2.5 times, thereby increasing the overall production efficiency of fissile material. Furthermore, as nuclear fission increases, the amount of heat generated and, therefore, the amount of power generated increases, and it becomes more reliable to secure in-house power consumption and provide surplus power.
本発明の目的は、核分裂性物質の生産効率及び
熱発生効率を高めた溶融塩核化学反応方法及びこ
の方法に使用するための改良された高利得型加速
器溶融塩増殖炉を提供することである。 An object of the present invention is to provide a molten salt nuclear chemical reaction method with increased fissile material production efficiency and heat generation efficiency, and an improved high-gain accelerator molten salt breeder reactor for use in this method. .
本願発明者等は、鋭意研究の結果、基本的には
前記sf AMSBにおいて、自ら生産した、または
供給を受けた核分裂性物質をターゲツト溶融塩中
に相当多量に混在させることによつてこの目的を
達し得ることを見出し、本願特許請求の範囲に記
載の本願発明の方法及び装置を達成した。 As a result of intensive research, the inventors of the present application basically achieved this objective by mixing a considerable amount of fissile material produced by themselves or supplied to the target molten salt in the sf AMSB. The inventors have found that it is possible to achieve this, and have achieved the method and apparatus of the present invention as set forth in the claims of the present application.
本願発明において使用するsf AMSBの構造は
基本的には前記先願発明のsf AMSBと同じであ
るが、ターゲツト塩総量を少くし、含有核分裂性
物質の絶対量を少くさせるように工夫されてお
り、またターゲツト溶融塩の組成については、こ
れに積極的に核分裂性物質を加える。目的ないし
設計法によつては0.2〜1mol％ 233UF4が考えら
れる。これによつて臨界になることはないが、さ
らに高濃度となると核分裂性物質の必要量が増大
して非経済性となる。この濃度は、生産された
233Uの供給を受けて稼動される予定の溶融塩発電
炉の燃料塩中の組成より充分に濃厚であるように
選んでこれをそのまま利用できるようにする。 The structure of the sf AMSB used in the present invention is basically the same as the sf AMSB of the prior invention, but it is devised to reduce the total amount of target salt and the absolute amount of fissile material contained. Furthermore, fissile material is actively added to the composition of the target molten salt. Depending on the purpose or design method, 0.2 to 1 mol% 233 UF 4 can be considered. Although this does not result in criticality, higher concentrations would increase the amount of fissile material required and make it uneconomical. This concentration was produced
The composition is chosen so that it is sufficiently richer than the composition of the fuel salt in the molten salt power reactor that is scheduled to be operated with the supply of 233 U, so that it can be used as is.
また、従来のAMSBにおいては、挾義の核破
砕反応と核分裂反応とが同程度に進行するが、本
願のごとき場合においては、核分裂反応が多くな
り、それだけ核反応廃棄物（不純物）が多くな
る。そのため必要であれば、本願発明の方法にお
いては、前記sf AMSBにおいて、ターゲツト溶
融塩バイパス系を設け、これに液体金属接触反応
装置及びニツケルウールフイルターを設置する。 In addition, in conventional AMSB, the nuclear spallation reaction and the nuclear fission reaction proceed at the same rate, but in the case of the present application, the number of nuclear fission reactions increases, and the amount of nuclear reaction waste (impurities) increases accordingly. . Therefore, if necessary, in the method of the present invention, a target molten salt bypass system is provided in the sf AMSB, and a liquid metal contact reaction device and a nickel wool filter are installed therein.
溶融塩は貯蔵タンクより適度（年間体積で10〜
30％程度）に汲み出し、精製操作を加えて溶融塩
発電炉に供給されるが、代りに塩組成の変動を修
正しながら核分裂性物質を含まないLiF−BeF2−
ThF4塩などを添加する。 Molten salt is more moderate than storage tanks (annual volume 10~
LiF−BeF 2 −, which does not contain fissile material while correcting fluctuations in salt composition, is pumped out to about 30%), subjected to refining operations, and then supplied to the molten salt power reactor.
Add ThF4 salt etc.
なお、本願発明においては 233UF4の代りに
235UF4または 239PuF3を用いて始動することもで
きる。また前記sf AMSBの出願明細書において
示したごときＵ混合塩をTh混合塩の代りに使用
し 239Pu生産炉として使用することもできる。 In addition, in the present invention, instead of 233 UF 4
It is also possible to start with 235 UF 4 or 239 PuF 3 . It is also possible to use the U mixed salt as shown in the specification of the SF AMSB application in place of the Th mixed salt and use it as a 239 Pu production furnace.
以下本発明を実施例によつて説明する。 The present invention will be explained below with reference to Examples.
実施例 １
sf AMSBにおいて、核反応容器は、図面に示
すように、塩溜りの上半分は直径約4.5mの円筒
状、下半分は逆円錐状となるごとく形状及び寸法
を若干変更したものを用いる。全体系の溶融塩容
量は約90m3であり、溶融塩組成としては、核分裂
性物質を添加した。Example 1 At the sf AMSB, the nuclear reaction vessel has a slightly modified shape and dimensions, with the upper half of the salt pool shaped like a cylinder with a diameter of approximately 4.5 m, and the lower half shaped like an inverted cone, as shown in the drawing. use The molten salt capacity of the entire system was approximately 90 m 3 , and fissile material was added to the molten salt composition.
7LiF−BeF2−ThF4− 233UF4（64−18−17.5−
0.5mol％）
を使用する。この場合、Th総量は約120ton、
233U必要量は約3.4tonである。 7LiF−BeF 2 −ThF 4 − 233 UF 4 (64−18−17.5−
0.5mol%) is used. In this case, the total amount of Th is approximately 120 tons,
The required amount of 233 U is approximately 3.4 tons.
これを使用したsf AMSBは核分裂性物質を添
加しない塩の場合より 233U生産効率が40〜50％
増大する。従つて、約1.2ton／年の率で核分裂性
物質が生産され、その倍増時間は２〜３年とな
る。これは年間約0.2mol％の増加を意味する。
よつて、組成がオーバーフロー配管１１を介して
ターゲツト溶融塩と平衡している貯蔵タンク１２
内の塩を月１回程度取り出し、核分裂性物質を含
まない塩7LiF−BeF2−ThF4（64−18−18mol％）
を同量添加する。 SF AMSB using this has a 233 U production efficiency of 40 to 50% compared to the case of salt without adding fissile material.
increase Therefore, fissile material is produced at a rate of about 1.2 tons/year, and the doubling time is 2 to 3 years. This means an increase of approximately 0.2 mol% per year.
Thus, the storage tank 12 whose composition is in equilibrium with the target molten salt via the overflow pipe 11
The salt inside is taken out about once a month, and the salt 7LiF−BeF 2 −ThF 4 (64−18−18 mol%) that does not contain fissile material is extracted.
Add the same amount.
核分裂による不純物等の量は、sf AMSBの場
合より増大する（年間約40Kgが約170Kgとなる）
ので、図面に示すごとく、溶融塩を連続精製する
ためのバイパス系２０を設ける。これは、熱交換
器出口配管７より分岐され、ポンプ２１により、
液体金属接触反応装置２２及びニツケルウールフ
イルター２３を経由して溶融塩主循環ポンプ９に
戻る系である。 The amount of impurities etc. due to nuclear fission will be larger than in the case of SF AMSB (approximately 40Kg per year becomes approximately 170Kg)
Therefore, as shown in the drawing, a bypass system 20 is provided for continuous purification of the molten salt. This is branched from the heat exchanger outlet pipe 7, and pumped by the pump 21.
This system returns to the molten salt main circulation pump 9 via the liquid metal contact reaction device 22 and the nickel wool filter 23.
貯蔵タンク１２から取り出されたターゲツト塩
はさらに精製されて、溶融塩発電炉用燃料塩の核
分裂性物質添加剤として利用される。また 233U
を抽出して、固体核燃料として利用してもよい。
この高利得型加速器溶融塩増殖炉の発電量は約33
万KW（熱）程度増大するが、これはターゲツト
溶融塩の流量を約１m3／秒増大させることにより
処理される。 The target salt removed from the storage tank 12 is further purified and utilized as a fissile material additive for fuel salt for molten salt power reactors. Also 233 U
may be extracted and used as solid nuclear fuel.
The power generation capacity of this high-gain accelerator molten salt breeder reactor is approximately 33
This increases the amount of heat by approximately 10,000 KW (heat), but this is handled by increasing the flow rate of the target molten salt by approximately 1 m 3 /sec.
実施例 ２
ターゲツト塩に添加する核分裂性物質は実施例
１の場合は 233Uであつたが、20％ 235U＋80％
238Uに置き換えることができる。この場合は約
14tonの238Uが約120tonの232Th中の14tonと置
換えられる。したがつて、また生産される 233U
中の約12％が 239Puとなる。しかし、その後
238U（および 235U）は追加されることがないの
で、次第に 239Puの濃度は減少する。また、この
核分裂性物質の供給を受けた溶融塩発電炉中で
は、一層急速に 239Puは消滅してゆく。Example 2 The fissile material added to the target salt was 233 U in Example 1, but 20% 235 U + 80%
Can be replaced with 238U. In this case approximately
14 tons of 238U will be replaced with 14 tons of approximately 120 tons of 232Th. Therefore, 233 U also produced
Approximately 12% of it is 239 Pu. But then
Since no 238 U (and 235 U) is added, the concentration of 239 Pu gradually decreases. Furthermore, in the molten salt power reactor supplied with this fissile material, 239 Pu disappears even more rapidly.
実施例 ３
初期添加の核分裂性物質を 239PuF3としてもほ
ぼ同様のもしくはより高い効率が期待できる。Example 3 Almost the same or higher efficiency can be expected even if 239 PuF 3 is used as the initially added fissile material.
以上からわかるように、反応の初期には核分裂
性物質を必要とするが、２〜４年以内の優れた倍
増時間が期待できる。また、核分裂性物質の生産
効率が50％以上増大させられる。 As can be seen from the above, although fissile material is required at the initial stage of the reaction, an excellent doubling time within 2 to 4 years can be expected. Additionally, the production efficiency of fissile material is increased by more than 50%.
さらに、発熱量が増加し、必要電力確保がより
確実となり電力売却の可能性もある。 Furthermore, the amount of heat generated will increase, making it more certain that the necessary power will be secured, and there is a possibility that the power will be sold.
本発明においては、ターゲツト溶融塩から取り
出した塩がそのまま溶融塩発電炉の燃料塩に投入
でき、核分裂性物質の添加に役立てることができ
る。ただし、一部の不純物除去処理は必要であ
る。 In the present invention, the salt extracted from the target molten salt can be directly put into the fuel salt of the molten salt power reactor, and can be used for adding fissile material. However, some impurity removal treatment is necessary.
本発明の装置は、設計変更を行つても新しいＲ
＆Ｄをほとんど要しない。従つて、発電炉仕様、
核エネルギー事情、経済性などを考慮しつつ、そ
の時点に最適の溶融塩組成による操業を可能なら
しめる。 Even if the device of the present invention undergoes a design change, the new R
&D is hardly required. Therefore, power reactor specifications,
While considering the nuclear energy situation and economic efficiency, it will be possible to operate with the optimal molten salt composition at that time.
図は本発明におけるsf AMSBの概要説明図で
ある。
図において、１……陽子ビーム（1GeV，
300mA）の入射孔、２……溶融塩の流入口、３
……流出通路、４……核化学反応装置、５……黒
鉛遮蔽体、６……黒鉛遮蔽体、７……配管、８…
…熱交換器、９……溶融塩循環ポンプ、１０……
流量調節弁、１１……オーバーフロー配管、１２
……溶融塩貯蔵タンク、１３……オリフイス、１
４……シヤター、１５……真空排気系、１６……
蒸気トラツプ、１７……線型加速器、１８……導
管、１９……陽子線集束用磁石、２０……塩精製
用バイパス配管、２１……バイパス用ポンプ、２
２……液体金属接触反応装置、２３……ニツケ
ル・ウール・フイルター。
The figure is a schematic explanatory diagram of the sf AMSB in the present invention. In the figure, 1...proton beam (1GeV,
300mA) incidence hole, 2...molten salt inlet, 3
... Outflow passageway, 4... Nuclear chemical reaction device, 5... Graphite shield, 6... Graphite shield, 7... Piping, 8...
...heat exchanger, 9...molten salt circulation pump, 10...
Flow rate control valve, 11... Overflow piping, 12
... Molten salt storage tank, 13 ... Orifice, 1
4... Shutter, 15... Vacuum exhaust system, 16...
Steam trap, 17... linear accelerator, 18... conduit, 19... proton beam focusing magnet, 20... bypass piping for salt purification, 21... bypass pump, 2
2... Liquid metal contact reactor, 23... Nickel wool filter.
Claims (1)
融塩核化学反応において、該溶融塩中に核分裂性
物質を含ませて反応を行うことを特徴とする溶融
塩核化学反応方法。 ２ 該核分裂性物質は 233UF4である第１項の方
法。 ３ 該核分裂性物質は 235UF4である第１項の方
法。 ４ 該核分裂性物質は 239PuF3である第１項の方
法。 ５ 溶融塩核反応容器、熱交換器及び溶融塩を該
核反応容器と該熱交換器との間を循環させるため
の溶融塩循環ポンプから成り、該核反応容器は内
壁に沿つて黒鉛遮蔽体を設けたハステロイＮから
作られた上下を閉じた円筒形の容器で上面に孔を
有し、その孔を通つて陽子などの高速荷電粒子が
容器内の溶融塩の液面に直接入射するように該容
器上に線型加速器を直結して設置した単一流体型
加速器溶融塩増殖炉において、液体金属接触反応
装置及びニツケルフイルターを設けたターゲツト
溶融塩バイパス系を設置したことを特徴とする溶
融塩核化学反応装置。[Scope of Claims] 1. A molten salt nuclear chemical reaction using a single-fluid accelerator molten salt breeder reactor, characterized in that the reaction is carried out by including a fissile material in the molten salt. Method. 2. The method of paragraph 1, wherein the fissile material is 233 UF 4 . 3. The method of paragraph 1, wherein the fissile material is 235 UF 4 . 4. The method of paragraph 1, wherein the fissile material is 239 PuF 3 . 5 consisting of a molten salt nuclear reaction vessel, a heat exchanger, and a molten salt circulation pump for circulating the molten salt between the nuclear reaction vessel and the heat exchanger, the nuclear reaction vessel having a graphite shield along its inner wall; It is a cylindrical container with a closed top and bottom made of Hastelloy N with a hole in the top surface, so that high-speed charged particles such as protons directly enter the liquid surface of the molten salt in the container through the hole. A molten salt core characterized in that a target molten salt bypass system equipped with a liquid metal contact reaction device and a nickel filter is installed in a single fluid type accelerator molten salt breeder reactor in which a linear accelerator is directly connected and installed on the vessel. Chemical reactor.
Priority Applications (4)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP56074897A JPS57190293A (en) | 1981-05-20 | 1981-05-20 | Melt salt nuclear reaction method and device |
| FR8208785A FR2519792A1 (en) | 1981-05-20 | 1982-05-19 | PROCESS AND APPARATUS FOR NUCLEAR CHEMICAL REACTION WITH MELT SALT |
| DE3218852A DE3218852A1 (en) | 1981-05-20 | 1982-05-19 | NUCLEAR MELTING MELT REACTION METHOD AND DEVICE FOR CARRYING OUT THIS METHOD |
| GB8214599A GB2098788A (en) | 1981-05-20 | 1982-05-19 | Process and apparatus for molten-salt nuclear chemical reaction process for circulating a fuel salt in a combination system of the apparatus with a molten- salt converter reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP56074897A JPS57190293A (en) | 1981-05-20 | 1981-05-20 | Melt salt nuclear reaction method and device |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS57190293A JPS57190293A (en) | 1982-11-22 |
| JPS6364754B2 true JPS6364754B2 (en) | 1988-12-13 |
Family
ID=13560635
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP56074897A Granted JPS57190293A (en) | 1981-05-20 | 1981-05-20 | Melt salt nuclear reaction method and device |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS57190293A (en) |
-
1981
- 1981-05-20 JP JP56074897A patent/JPS57190293A/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS57190293A (en) | 1982-11-22 |
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