JPS5896277A - Method of circulating fuel of accelerator melted salt breeder reactor - Google Patents
Method of circulating fuel of accelerator melted salt breeder reactorInfo
- Publication number
- JPS5896277A JPS5896277A JP56194443A JP19444381A JPS5896277A JP S5896277 A JPS5896277 A JP S5896277A JP 56194443 A JP56194443 A JP 56194443A JP 19444381 A JP19444381 A JP 19444381A JP S5896277 A JPS5896277 A JP S5896277A
- Authority
- JP
- Japan
- Prior art keywords
- salt
- fuel
- reactor
- breeder reactor
- target
- 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
- Y02E30/30—Nuclear fission reactors
Landscapes
- Solid Fuels And Fuel-Associated Substances (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】 本発明は加速器溶融塩増殖炉の燃料循環方法に関する。[Detailed description of the invention] The present invention relates to a fuel circulation method for an accelerator molten salt breeder reactor.
詳しくは高利得式加速器溶融塩増殖炉（ｈ、ｇ、ＡＭ８
Ｂと略称する）と溶融塩発電炉（ＭＳＯＲと略称する）
との組合せ系における燃料塩の循環方法に関する。For details, please refer to the high-gain accelerator molten salt breeder reactor (h, g, AM8
(abbreviated as B) and molten salt power reactor (abbreviated as MSOR)
This invention relates to a method for circulating fuel salt in a combination system.
本願発明者は先にｈ、ｇ、ＡＭ８Ｂを発明し発明の名称
、溶融塩核化学反応方法および装置〃として特許出願（
特願昭５６−７４８９７）し、その明細書中にｈ、ｇ、
ＡＭＯＢから汲み出されたターゲツト溶融塩がＭＳＯＲ
の燃料塩に投入できることを開示した。The inventor of this application previously invented h, g, and AM8B and filed a patent application with the title of the invention as molten salt nuclear chemical reaction method and apparatus (
(Japanese Patent Application No. 56-74897), and in the specification h, g,
The target molten salt pumped from AMOB is MSOR
The company disclosed that it can be put into fuel salt.
本発明の目的はｈ、ｇ、ＡＭＳＲで生産された核分裂性
物質２ｍ３Ｕｖｉ−ＭＳＯＲに供給するための最も合理
化された核燃料循環方法を提供することである。The purpose of the present invention is to provide the most streamlined nuclear fuel circulation method for supplying 2m3 Uvi-MSOR of fissile material produced in h, g, AMSR.
本願発明者は、この目的を達成するため鋭意研究の結果
、ｈ、ｇ、ＡＭｓＢとＭＳＯＲとの組合せ系において、
０．２〜１ｍｏｌ％の””ＵＦ４を含む該ＡＭＳＢから
のターゲツト塩を該ＭＳＯＲの燃料塩に添加してＨ３Ｕ
ｐ４を補給するＶＣあたり、該ＭＳＯＲから汲み出され
た燃料塩を再処理して取り出された２３３Ｕｐ４を該タ
ーゲツト塩に添加して該ターゲツト塩の１３３ＵＩＦ、
濃度を高めることから成るｈ、ｇ、ＡＭＳＢの燃料塩の
循環方法を発明するに至つた。As a result of intensive research to achieve this objective, the inventor of the present application has found that in a combination system of h, g, AMsB and MSOR,
Target salt from the AMSB containing 0.2-1 mol% UF4 was added to the fuel salt of the MSOR to generate H3U.
For each VC that replenishes p4, 233Up4 extracted by reprocessing the fuel salt pumped from the MSOR is added to the target salt to increase the 133UIF of the target salt.
We have now devised a method for recycling h, g, and AMSB fuel salts, which consists of increasing their concentration.
而して、ＡＭＳＢから汲み出されたターゲツト塩はＭ８
０Ｒに輸送される前に酸化物除去操作、濾過による固形
懸濁物の除去などを含む精製操作を行うことが必要であ
る。Therefore, the target salt pumped from AMSB is M8
It is necessary to carry out purification operations including oxide removal operations, removal of solid suspensions by filtration, etc. before being transported to the OR.
ＭＳＯＲから取り出された燃料塩はＡＭＳＢの所在する
地域センター（監視された管理区域）内の再処理施設に
持ち込み、まづＨ！ＵＦ４のみを取り出す。The fuel salt extracted from the MSOR is brought to the reprocessing facility within the regional center (monitored management area) where AMSB is located, and the fuel salt is transported to the reprocessing facility within the AMSB's regional center (monitored control area). Take out only UF4.
それにはＦ２ガスによるフツ化処理で２３３ＵＦ、揮発
法などを利用し取り出された２ｍ３Ｕｐ、をＵＦ、に還
元しつつ補給用ターゲツト塩に添加する。また、ＭＳＣ
！Ｈの燃料塩組成を等しくするため’ＬｉＦ、ＢｅＦ＊
の添加が適宜性はれる。To do this, 233 UF is removed through fluorination treatment with F2 gas, and 2 m3 Up is extracted using a volatilization method, etc., which is reduced to UF and added to the target salt for replenishment. Also, MSC
! In order to equalize the fuel salt composition of H'LiF, BeF*
It is advisable to add .
再処理されてクリーンになつたＴｈＦ４含有塩はＴｈＦ
４及び少量のＢｅＦ２などを添加してターゲツト塩とほ
ぼ同組成の稀釈用塩に調整してＡＭＳＢに随時に添加す
る。ThF4-containing salt that has been reprocessed and becomes clean is ThF
4 and a small amount of BeF2, etc. to prepare a diluent salt having approximately the same composition as the target salt, and add it to AMSB as needed.
次に実施例によつて本発明を具体的に説明する。Next, the present invention will be specifically explained with reference to Examples.
実施例
ＩＧｅＶ、３００ｍＡ陽子加速器を有しターゲツト塩と
して’ＬｉＦＢｅＦｚ−ＴｈＦ４（！”ＩＵＦｉ、（６
４−１８１７，５０−５ｍｏ１％）を用いたｈ、ｇ、Ａ
ＭＳＢは年間約１，１３ｔｏｎの２３３Ｕ燃料を生産す
ることかできる。したがつて、１００万ＫＷｅのＭＳＯ
Ｒ（転換率０．９）の１７基分に必要な核分裂性物質を
供給できる。Example IGeV with 300 mA proton accelerator and 'LiFBeFz-ThF4(!'IUFi, (6
h, g, A using 4-1817, 50-5 mo1%)
MSB can produce approximately 1.13 tons of 233U fuel per year. Therefore, an MSO of 1 million KWe
It is possible to supply the fissile material necessary for 17 units of R (conversion rate 0.9).
以下、ＭＳＯＲ１基を対象として燃料サイクルを説明す
る。Hereinafter, the fuel cycle will be explained for one MSOR unit.
ＭＳＯＲから取り出された６５０ｋｇの塩は’ＬｉＦ−
ＢｅＦ’５−ＴｈＦ４」”Ｕｆ’４（７１，７５１６１
２０，２５ｍｏｌ％）の組成であり、地域センターの再
処理施設で約７．８−の！１ｍＵｐ４がフツ化物揮発法
で分離される。650 kg of salt removed from MSOR is 'LiF-
BeF'5-ThF4""Uf'4(71,75161
20.25 mol %), and the reprocessing facility at the regional center has a composition of approximately 7.8-! 1 mUp4 is isolated by fluoride volatilization method.
これをＡＭＳＢから取り出されたターゲツト塩５６０ｋ
ｇに添加すると’ＬｉＦＢｅＦ２ＴｈＦ４”３ＵＦ４（
６３，８１７，９１７，４０，８６ｍｏ１％）となる。This is target salt 560k extracted from AMSB.
When added to 'LiFBeF2ThF4''3UF4 (
63,817,917,40,86mo1%).
これは・Ｍ８０Ｒから取り出された量を差引くと約８ｋ
ｆの２３３Ｕを添加したことを意味する。This is approximately 8k after subtracting the amount taken out from M80R.
This means that 233U of f was added.
このような操作を年間約１ｏ回行うこととする。It is assumed that such operations are performed approximately 10 times a year.
なお、Ｍ５ＯＲに添加する際、組成を調整するため’Ｌ
ｉＦ約７５に４とＢｅＦ２約１８ｋｇを同時に添加する
。In addition, when adding to M5OR, 'L' is added to adjust the composition.
4 and approximately 18 kg of BeF2 are added to iF approximately 75 at the same time.
本発明は、最小限の操作で最も経済的な燃料サイクルを
構成でき、また燃料塩がフツ化物であるため化学的に安
全であり、揮発性廃棄物もＡＭＳＢまたはＭ８０Ｒで処
理されており、また最小限の物量輸送で済むなどの効果
を有する。The present invention allows the most economical fuel cycle to be constructed with a minimum of operations, is chemically safe because the fuel salt is a fluoride, volatile wastes are also treated with AMSB or M80R, and This has the advantage of requiring only a minimum amount of material to be transported.
さらにまた、本発明の方法においては、監視された地区
センター以外には、核分裂性物質は放射性物質をかなり
多量に含んだ溶融塩を凝固させた物質として輸送される
ので、高いｒ放射能を有し簡単に核分裂性物質を分離す
ることはできない。Furthermore, in the method of the present invention, the fissile material is transported to areas other than the monitored district center as a solidified material of molten salt containing a considerable amount of radioactive material, and therefore has a high r-radioactivity. However, fissile material cannot be easily separated.
しかも２３２Ｕ、２３４Ｕ（２３５Ｕ）、２３６Ｕを含
有して２３３Ｕと複雑な同位元素構成を有し、臨界量は
大きく容易に爆発物とすることができない効果を有する
。In addition, it contains 232U, 234U (235U), and 236U, and has a complex isotope composition with 233U, and has a large critical mass that makes it difficult to make it into an explosive.
図は本発明の燃料循環方法の１具体例プロセスを示す。 図において、ＩＦｌＩＧｅＶ３００ｍＡ陽子加速器。 ２はｈ．ｇ．ＡＭＳＢ ３はタンク ４は再処理施設 ５はＭＳＣＲ ６はタンク The figure shows one example process of the fuel circulation method of the present invention. In the figure, IFlIGeV 300mA proton accelerator. 2 is h. g. AMSB 3 is a tank 4 is a reprocessing facility 5 is MSCR 6 is a tank
Claims (1)
合せ系において、０．７〜１ｍｏｌ％の２３３０Ｆ、を
含む該増殖炉からのターゲツト塩を該発電炉の燃料塩に
添加して２３３ＵＦ４を補給するにあたり、該発電炉か
ら汲み出された燃料塩を再処理して取り出さ扛た２３３
ＵＦ４を、該ターゲツト塩に添加して該ターゲツト塩の
２３３ＵＦ４濃度を高めることから成る加速器溶融塩増
殖炉の燃料循環方法。 ２、溶融塩の組成は’ＬｉＦＢｅＦ２ＴｈＦ４−２３３
ＵＦ４である第１項の燃料循環方法。 ３、該増殖炉からのターゲツト塩は不純物除去の操作が
加えられる第１項の方法。 ４、該発電炉から汲み出された燃料塩は再処理施設にお
いてフツ化ガス処理を行つて２３３ＵＦ４を２３３ＵＦ
・として取り出しこれを還元なして２３３ＵＦ４とする
第１項の方法。 ５、再処理施設でクリーンになつたＴｈＦ４含有塩はＴ
ｈＦ４及び少量のＢｅＦ２を添加してターゲツト塩とほ
ぼ同組成の稀釈用塩に調整して該増殖炉に適時に添加さ
れる第１項の方法。[Claims] l. In a combination system of a high-gain accelerator molten salt breeder reactor and a molten salt power reactor, target salt from the breeder reactor containing 0.7 to 1 mol% 2330F is used in the power reactor. When adding 233UF4 to the fuel salt, the fuel salt pumped out from the power reactor was reprocessed and extracted.
A method of fuel circulation in an accelerator molten salt breeder reactor comprising adding UF4 to the target salt to increase the 233UF4 concentration of the target salt. 2. The composition of the molten salt is 'LiFBeF2ThF4-233
The fuel circulation method of item 1 which is UF4. 3. The method of item 1, wherein the target salt from the breeder reactor is subjected to an operation for removing impurities. 4. The fuel salt pumped out from the power reactor is treated with fluoride gas at a reprocessing facility to convert 233UF4 to 233UF.
・The method of item 1, which is extracted as 233UF4 without reduction. 5. The ThF4-containing salt cleaned at the reprocessing facility is T
1. The method of item 1, wherein hF4 and a small amount of BeF2 are added to prepare a diluent salt having approximately the same composition as the target salt, and the diluent salt is added to the breeder reactor in a timely manner.
Priority Applications (4)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP56194443A JPS5896277A (en) | 1981-12-04 | 1981-12-04 | Method of circulating fuel of accelerator melted salt breeder reactor |
| 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 |
|---|---|---|---|
| JP56194443A JPS5896277A (en) | 1981-12-04 | 1981-12-04 | Method of circulating fuel of accelerator melted salt breeder reactor |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS5896277A true JPS5896277A (en) | 1983-06-08 |
Family
ID=16324661
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP56194443A Pending JPS5896277A (en) | 1981-05-20 | 1981-12-04 | Method of circulating fuel of accelerator melted salt breeder reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS5896277A (en) |
-
1981
- 1981-12-04 JP JP56194443A patent/JPS5896277A/en active Pending
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