JPS5847040B2 - Plutonium separation method - Google Patents
Plutonium separation methodInfo
- Publication number
- JPS5847040B2 JPS5847040B2 JP53041052A JP4105278A JPS5847040B2 JP S5847040 B2 JPS5847040 B2 JP S5847040B2 JP 53041052 A JP53041052 A JP 53041052A JP 4105278 A JP4105278 A JP 4105278A JP S5847040 B2 JPS5847040 B2 JP S5847040B2
- Authority
- JP
- Japan
- Prior art keywords
- plutonium
- uranium
- nitric acid
- concentration
- lactic acid
- 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
- 229910052778 Plutonium Inorganic materials 0.000 title claims description 61
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 title claims description 61
- 238000000926 separation method Methods 0.000 title description 21
- JVTAAEKCZFNVCJ-UHFFFAOYSA-N lactic acid Chemical compound CC(O)C(O)=O JVTAAEKCZFNVCJ-UHFFFAOYSA-N 0.000 claims description 44
- 229910052770 Uranium Inorganic materials 0.000 claims description 36
- 238000000034 method Methods 0.000 claims description 24
- 229910017604 nitric acid Inorganic materials 0.000 claims description 24
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims description 23
- 235000014655 lactic acid Nutrition 0.000 claims description 22
- 239000004310 lactic acid Substances 0.000 claims description 22
- 239000000243 solution Substances 0.000 claims description 13
- 239000011259 mixed solution Substances 0.000 claims description 8
- 239000003795 chemical substances by application Substances 0.000 claims description 7
- 239000002904 solvent Substances 0.000 claims description 6
- 229910002651 NO3 Inorganic materials 0.000 claims description 4
- NHNBFGGVMKEFGY-UHFFFAOYSA-N Nitrate Chemical compound [O-][N+]([O-])=O NHNBFGGVMKEFGY-UHFFFAOYSA-N 0.000 claims description 4
- 125000005289 uranyl group Chemical group 0.000 claims description 4
- 229910002007 uranyl nitrate Inorganic materials 0.000 claims description 4
- OAICVXFJPJFONN-UHFFFAOYSA-N Phosphorus Chemical compound [P] OAICVXFJPJFONN-UHFFFAOYSA-N 0.000 claims description 3
- 229910052698 phosphorus Inorganic materials 0.000 claims description 3
- 239000011574 phosphorus Substances 0.000 claims description 3
- ZQPKENGPMDNVKK-UHFFFAOYSA-N nitric acid;plutonium Chemical compound [Pu].O[N+]([O-])=O ZQPKENGPMDNVKK-UHFFFAOYSA-N 0.000 claims description 2
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 claims 1
- DNYWZCXLKNTFFI-UHFFFAOYSA-N uranium Chemical compound [U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U][U] DNYWZCXLKNTFFI-UHFFFAOYSA-N 0.000 description 35
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 14
- 239000003638 chemical reducing agent Substances 0.000 description 8
- 238000000605 extraction Methods 0.000 description 6
- 238000012958 reprocessing Methods 0.000 description 5
- 238000012360 testing method Methods 0.000 description 5
- AFCIMSXHQSIHQW-UHFFFAOYSA-N [O].[P] Chemical compound [O].[P] AFCIMSXHQSIHQW-UHFFFAOYSA-N 0.000 description 4
- 239000002253 acid Substances 0.000 description 4
- 239000007864 aqueous solution Substances 0.000 description 4
- 239000007788 liquid Substances 0.000 description 4
- BDAGIHXWWSANSR-UHFFFAOYSA-N methanoic acid Natural products OC=O BDAGIHXWWSANSR-UHFFFAOYSA-N 0.000 description 4
- 238000006722 reduction reaction Methods 0.000 description 4
- SMCVJROYJICEKJ-UHFFFAOYSA-N 2-hydroxypropanoic acid;nitric acid Chemical compound O[N+]([O-])=O.CC(O)C(O)=O SMCVJROYJICEKJ-UHFFFAOYSA-N 0.000 description 3
- WYKYKTKDBLFHCY-UHFFFAOYSA-N chloridazon Chemical compound O=C1C(Cl)=C(N)C=NN1C1=CC=CC=C1 WYKYKTKDBLFHCY-UHFFFAOYSA-N 0.000 description 3
- 238000010586 diagram Methods 0.000 description 3
- 150000002148 esters Chemical class 0.000 description 3
- 238000011084 recovery Methods 0.000 description 3
- 239000002915 spent fuel radioactive waste Substances 0.000 description 3
- -1 uranium-hydrazine Chemical compound 0.000 description 3
- OSWFIVFLDKOXQC-UHFFFAOYSA-N 4-(3-methoxyphenyl)aniline Chemical compound COC1=CC=CC(C=2C=CC(N)=CC=2)=C1 OSWFIVFLDKOXQC-UHFFFAOYSA-N 0.000 description 2
- 238000006243 chemical reaction Methods 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- 235000019253 formic acid Nutrition 0.000 description 2
- OAKJQQAXSVQMHS-UHFFFAOYSA-N hydrazine Substances NN OAKJQQAXSVQMHS-UHFFFAOYSA-N 0.000 description 2
- 239000000463 material Substances 0.000 description 2
- 239000003758 nuclear fuel Substances 0.000 description 2
- 238000011160 research Methods 0.000 description 2
- 239000000126 substance Substances 0.000 description 2
- 239000002699 waste material Substances 0.000 description 2
- JPGXOMADPRULAC-UHFFFAOYSA-N 1-[butoxy(butyl)phosphoryl]oxybutane Chemical compound CCCCOP(=O)(CCCC)OCCCC JPGXOMADPRULAC-UHFFFAOYSA-N 0.000 description 1
- 229910052695 Americium Inorganic materials 0.000 description 1
- AVXURJPOCDRRFD-UHFFFAOYSA-N Hydroxylamine Chemical compound ON AVXURJPOCDRRFD-UHFFFAOYSA-N 0.000 description 1
- LXQXZNRPTYVCNG-UHFFFAOYSA-N americium atom Chemical compound [Am] LXQXZNRPTYVCNG-UHFFFAOYSA-N 0.000 description 1
- 238000013459 approach Methods 0.000 description 1
- 239000007853 buffer solution Substances 0.000 description 1
- 230000003139 buffering effect Effects 0.000 description 1
- 238000001311 chemical methods and process Methods 0.000 description 1
- 238000004140 cleaning Methods 0.000 description 1
- 238000007796 conventional method Methods 0.000 description 1
- 230000007423 decrease Effects 0.000 description 1
- 230000002542 deteriorative effect Effects 0.000 description 1
- 230000007062 hydrolysis Effects 0.000 description 1
- 238000006460 hydrolysis reaction Methods 0.000 description 1
- 239000012535 impurity Substances 0.000 description 1
- 238000012986 modification Methods 0.000 description 1
- 230000004048 modification Effects 0.000 description 1
- 238000005192 partition Methods 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 229910052761 rare earth metal Inorganic materials 0.000 description 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 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
-
- 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
- Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
- Y02W30/00—Technologies for solid waste management
- Y02W30/50—Reuse, recycling or recovery technologies
Landscapes
- Extraction Or Liquid Replacement (AREA)
- Inorganic Compounds Of Heavy Metals (AREA)
Description
【発明の詳細な説明】
本発明は、使用済核燃料の再処理化学工程、再処理廃液
の処理工程でとりわけ有効なプルトニウムの分離方法に
関し、更に詳しくは燐の酸素酸エステル中にウランと共
存するプルトニウムを、その原子価を変えることなく乳
酸を用いることによって分離する方法に関するものであ
る。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a method for separating plutonium which is particularly effective in the chemical process of reprocessing spent nuclear fuel and the process of treating reprocessing waste liquid, and more specifically relates to a method for separating plutonium that coexists with uranium in a phosphorus oxygen acid ester. This invention relates to a method for separating plutonium using lactic acid without changing its valence.
現在、核燃料物質の再処理法として様々な方法が提案さ
れているが、実用化されているものとしてビューレック
ス湿式法がある。Currently, various methods have been proposed as methods for reprocessing nuclear fuel materials, and one that has been put into practical use is the Burex wet method.
この主要な目的は、使用済核燃料中に含まれているウラ
ンやプルトニウムを核分裂生成物から分離すること、ウ
ランとプルトニウムを分離し単離することにある。The main purpose of this is to separate the uranium and plutonium contained in spent nuclear fuel from the fission products, and to separate and isolate uranium and plutonium.
核分裂生成分とウランやプルトニウムとの分離には、３
０％に希釈された燐酸トリブチル（以下、「ＴＢＰＪと
略記する）が抽出剤として用いられ、ウランやプルトニ
ウムをＴＢＰに選択的に抽出する方法がとられている。Separation of fission products from uranium and plutonium requires 3
Tributyl phosphate diluted to 0% (hereinafter abbreviated as "TBPJ") is used as an extractant to selectively extract uranium and plutonium into TBP.
次に、ウランとプルトニウムとを分離するには、ウラナ
ス−ヒドラジン、第１鉄−ヒドラジン、あるいはヒドロ
キシルアミンのようなプルトニウム還元剤が用いられ、
ＴＢＰに抽出されているウラン、プルトニウムをこのよ
うな還元剤と接触させることによって、プルトニウムの
みＴＢＰに抽出されない原子価、つまり３価に還元し、
ウランとプルトニウムとを分離している。A plutonium reducing agent such as uranium-hydrazine, ferrous-hydrazine, or hydroxylamine is then used to separate the uranium and plutonium.
By bringing the uranium and plutonium extracted into TBP into contact with such a reducing agent, only plutonium is reduced to a valence that is not extracted into TBP, that is, trivalent.
Uranium and plutonium are separated.
このようにプルトニウムの分離法として原子価を調整す
る方法は非常に効率が良いが、対象とするプルトニウム
の量が増加するにつれて必要とする還元剤の量も増加す
る欠点がある。Although this method of separating plutonium by adjusting the valence is very efficient, it has the disadvantage that as the amount of target plutonium increases, the amount of reducing agent required also increases.
すなわち、前述したウラナス以外の還元剤を用いた場合
には、プルトニウムの純度を低下させたり、処理によっ
て発生する廃棄物の量を増加させたりするため、必ずし
も好ましい方法ではないのである。That is, if a reducing agent other than the above-mentioned uranus is used, it is not necessarily a preferable method because it lowers the purity of plutonium and increases the amount of waste generated during treatment.
そこで、現在のビューレックス湿式再処理工程において
は、ウラナスによる還元法が主として採用されている。Therefore, in the current Burex wet reprocessing process, the reduction method using Uranus is mainly adopted.
ところが、ウラナスのような核燃料物質を還元剤とした
場合には、工程内に溜留するウランの増加につながり得
策ではない。However, using a nuclear fuel material such as uranus as a reducing agent is not a good idea as it increases the amount of uranium that accumulates in the process.
しかも将来、高速増殖炉等から発生する使用済燃料を再
処理しようとする場合、含まれているプルトニウムが大
量なため、必要とするウラナスの量が工程処理能力をこ
えると考えられているから、在来工程のままではプルト
ニウムの処理が困難となってくる。Moreover, in the future, if spent fuel generated from fast breeder reactors, etc. is to be reprocessed, it is thought that the amount of uranus required will exceed the processing capacity due to the large amount of plutonium it contains. It will be difficult to process plutonium using conventional processes.
また現在、研究室的な段階では、前記のような欠点のな
いプルトニウムの還元法として電解による方法が西ドイ
ツのカールスルウェー研究所等で行われている。Currently, at the laboratory stage, an electrolytic method for reducing plutonium, which does not have the above-mentioned drawbacks, is being carried out at the Karlsruwe Institute in West Germany.
しかし、この電解法では特殊な電極をもつ抽出器を必要
とし、更に抽出器内の各段で精密な電解電流のコントロ
ールなどを要求されるなどの困難な因子を含んでいる。However, this electrolytic method requires an extractor with special electrodes, and also includes difficult factors such as requiring precise control of the electrolytic current at each stage within the extractor.
更に、プルトニウムの分離に還元反応を伴わない方法と
して蟻酸を用いた方法などがフランスにおいて試験され
ているが、逆抽出に用いる蟻酸の濃度は２モル程度の高
濃度を要し、ウランとの分離に最も重要である分離係数
は３程度で非常に低く、このため、さらに分離係数を上
げる条件を見出す必要がある。Furthermore, a method using formic acid that does not involve a reduction reaction to separate plutonium is being tested in France, but the concentration of formic acid used for back extraction requires a high concentration of about 2 molar, making it difficult to separate it from uranium. The separation factor, which is the most important factor in this, is very low at around 3, and therefore it is necessary to find conditions for further increasing the separation factor.
本発明の目的は、上記のような従来の分離法の欠点を解
消し、還元剤の添加や電気的還元操作を伴うことなく、
抽出操作のみによって容易にプルトニウムの分離を行え
、しかも分離性能もウラナス法と同じ位に良好な方法を
提供することにある。The purpose of the present invention is to eliminate the drawbacks of the conventional separation methods as described above, and to eliminate the need for adding a reducing agent or electrical reduction operation.
The object of the present invention is to provide a method that allows plutonium to be easily separated using only an extraction operation, and whose separation performance is as good as that of the Uranus method.
即ち本発明は、硝酸ウラニルＵＯ２（ＮＯａ ）２と共
に硝酸プルトニウムＰ ｕ （ＮＯ３）４を含ム燐の酸
素酸エステルを溶媒とする溶液を、乳酸水溶液と低濃度
の硝酸との混合水溶液からなる分離剤と接触させること
により、硝酸ウラニルを含有する燐の酸素酸を溶媒とす
る溶液と、プルトニウムを含有する前記分離剤を溶媒と
する溶液とに分離することを特徴とするウランとプルト
ニウムとを分離する方法である。That is, the present invention separates a solution containing plutonium nitrate P u (NO3)4 together with uranyl nitrate UO2(NOa)2 using a mixed aqueous solution of a lactic acid aqueous solution and a low concentration nitric acid as a solvent. Separation of uranium and plutonium, characterized by separating uranium and plutonium into a solution containing uranyl nitrate using a phosphorus oxyacid as a solvent and a solution containing plutonium using the separating agent as a solvent. This is the way to do it.
従って、本発明においては、プルトニウムは、その原子
価を変えることなくウランから分離される。Therefore, in the present invention, plutonium is separated from uranium without changing its valence.
分離剤としては乳酸水溶液と低濃度の硝酸との混合水溶
液が用いられる。As the separating agent, a mixed aqueous solution of lactic acid aqueous solution and low concentration nitric acid is used.
また、燐の酸素酸エステルとしてはＴＢＰのほかジブチ
ルブチルホスホネート（以下、「ＤＢＢＰ」と略記する
）等がある。In addition to TBP, examples of phosphorus oxygen acid esters include dibutylbutylphosphonate (hereinafter abbreviated as "DBBP") and the like.
乳酸はｐＨの緩衝溶液として一般に用いられている。Lactic acid is commonly used as a pH buffer solution.
例えば、米国のオークリッジ研究所や西ドイツのカール
スルウェー研究所等ではアメリシウム−希土類元素の分
離においてｐＨ条件を一定に維持するための緩衝効果と
合せて約１モルの高濃度で使用されているが、本発明と
は技術思想が明らかに異なるものである。For example, in the Oak Ridge Research Institute in the United States and the Carl Sulway Research Institute in West Germany, it is used at a high concentration of about 1 molar in the separation of americium and rare earth elements, along with a buffering effect to maintain a constant pH condition. , the technical idea is clearly different from the present invention.
以下本発明について、燐の酸素酸としてＴＢＰを用いた
場合を例にとって更に詳しく説明する。The present invention will be described in more detail below, taking as an example the case where TBP is used as the phosphorus oxygen acid.
各濃度の硝酸−乳酸の混合溶液とＴＢＰとの間でウラン
またはプルトニウムを分配させたとき、硝酸−乳酸混合
溶液中におけるウランまたはプルトニウムの濃度に対す
るＴＢＰ中におけるウランまたはプルトニウムの濃度の
比、すなわち分配係数は：硝酸濃度を一定とした場合、
共存する乳酸濃度が高くなるほど低下する。When uranium or plutonium is distributed between a mixed solution of nitric acid and lactic acid at various concentrations and TBP, the ratio of the concentration of uranium or plutonium in TBP to the concentration of uranium or plutonium in the mixed solution of nitric acid and lactic acid, that is, the distribution The coefficient is: When the nitric acid concentration is constant,
It decreases as the coexisting lactic acid concentration increases.
一方、乳酸濃度を一定とした場合、共存する硝酸濃度が
増加するに従いそれぞれの分配係数は高くなり、約２規
定以上の硝酸濃度に達するとほとんど乳酸の影響は無く
なり、単なる硝酸溶液中での分配係数に接近することが
試験の結果判明している。On the other hand, when the lactic acid concentration is constant, the distribution coefficients increase as the coexisting nitric acid concentration increases, and when the nitric acid concentration reaches approximately 2N or higher, the influence of lactic acid disappears, and the distribution coefficient in a simple nitric acid solution increases. Test results have shown that the coefficient approaches the
この中で、硝酸濃度が０．５規定、乳酸濃度が０．３モ
ル程度の混合溶液のＴＢＰに対するウラン、プルトニウ
ムの分配係数はウランで約３、プルトニウムで０．０７
程度であり、特にプルトニウムについては乳酸の効果が
著しく表われている。Among these, the distribution coefficients of uranium and plutonium to TBP for a mixed solution with a nitric acid concentration of 0.5N and a lactic acid concentration of about 0.3M are approximately 3 for uranium and 0.07 for plutonium.
The effect of lactic acid is particularly noticeable for plutonium.
従って、ＴＢＰ中に抽出されているウラン・プルトニウ
ムは低濃度の硝酸−乳酸の混合溶液により原子価を変え
ることなく容易に分離することが可能である。Therefore, the uranium and plutonium extracted in TBP can be easily separated using a mixed solution of nitric acid and lactic acid at a low concentration without changing the valence.
乳酸を含む溶液として回収されたプルトニウムは、その
溶液中の硝酸濃度を高める（約２規定）ことによって容
易にＴＢＰやＤＢＢＰに抽出され、乳酸と分離すること
ができる。Plutonium recovered as a solution containing lactic acid can be easily extracted into TBP and DBBP and separated from lactic acid by increasing the nitric acid concentration in the solution (about 2N).
その後、プルトニウムは、従来と同様の方法でＴＢＰや
ＤＢＢＰか−ら分離すればよい。Thereafter, plutonium may be separated from TBP and DBBP using a conventional method.
つまり、このような乳酸−硝酸溶液中におけるプルトニ
ウムの反応は、各硝酸濃度において、次のような化学形
をとる。That is, the reaction of plutonium in such a lactic acid-nitric acid solution takes the following chemical form at each nitric acid concentration.
硝酸０．５規定以下の場合
Ｐｕ（Ｌａｃ、）ｎ＞Ｐｕ（ＮＯｓ）＋ ”−（１
）硝酸２規定以上の場合
Ｐｕ（Ｌａｃ、）ｎ＜Ｐｕ（ＮＯ３）４ ・＝・（
２）但し、Ｌａｃ、は乳酸を示す。When nitric acid is 0.5N or less, Pu(Lac,)n>Pu(NOs)+ ”-(1
) When nitric acid is 2N or higher, Pu(Lac,)n<Pu(NO3)4 ・=・(
2) However, Lac represents lactic acid.
しかし、ウランについては、いずれの硝酸濃度において
も次のようになる。However, for uranium, the following is true at any nitric acid concentration.
ＵＯ２（Ｌ ａ ｃ 、 ） ｎ（＜ＵＯ２（ＮＯ３）
２ＴＢＰやＤＢＢＰに抽出されるウラン、プルトニウム
は、次式に示されるように、硝酸化合物を形成して抽出
される。UO2(L a c , ) n(<UO2(NO3)
Uranium and plutonium extracted into 2TBP and DBBP are extracted by forming a nitric acid compound as shown in the following formula.
Ｍ＋ｎＮｏ３−十ｍ０ＲＧ＃ｇ（Ｎｏ３）ｎ ・ｍ０Ｒ
Ｇ但し、Ｍ：ウラン、プルトニウム
ＯＲＧ：燐の酸素酸エステル
従って、前記（１） 、 （２）式に示したような反応
が起こる条件に硝酸濃度を調整すれば、抽出や逆抽を自
由に行うことができ、ウランからの分離、さらには分離
剤として用いた乳酸からのプルトニウムの分離も容易に
行うことができる。M+nNo3-10m0RG#g(No3)n ・m0R
G However, M: uranium, plutonium ORG: phosphorus oxygen acid ester Therefore, if the nitric acid concentration is adjusted to the conditions where the reactions shown in equations (1) and (2) above occur, extraction and back extraction can be performed freely. plutonium can be easily separated from uranium, and furthermore, plutonium can be easily separated from lactic acid used as a separating agent.
乳酸によってプルトニウムを分離回収するための全体の
流れ図を示せば第１図の如くである。The overall flowchart for separating and recovering plutonium using lactic acid is shown in Figure 1.
なお、ここで低濃度の硝酸とは約２規定未満の硝酸を意
味する。Note that the term "low concentration nitric acid" as used herein means nitric acid having a concentration of less than about 2N.
０．５規定であれば前記のように分離効率は良好である
が、それ以上、あるいはそれ以下でもウランとプルトニ
ウムとの分配係数に差があればミキサセトラの段数を変
えることなどにより分離が可能だからである。If the standard is 0.5, the separation efficiency is good as mentioned above, but if there is a difference in the distribution coefficient between uranium and plutonium even if it is higher or lower, separation can be achieved by changing the number of stages of the mixer settler. It is.
また、乳酸濃度は、たとえ高くてもウランとプルトニウ
ムとの分配係数に差があるから分離可能であるが、系に
異物（不純物）が本＊入ることは好ましくないし、効率
の点からしても乳酸濃度は低い方が好ましい。Furthermore, even if the concentration of lactic acid is high, it can be separated because there is a difference in the partition coefficient between uranium and plutonium, but it is undesirable for foreign substances (impurities) to enter the system, and from the point of view of efficiency. A lower lactic acid concentration is preferable.
次に、本発明の実施例について説明する。Next, examples of the present invention will be described.
実施例 ウラン・プルトニウム分離 ウラン・プルトニウム分離工程を第２図に示す。Example Uranium/plutonium separation Figure 2 shows the uranium/plutonium separation process.
この工程は、現在用いられている軽水炉型ビューレック
ス湿式再処理工程において、プルトニウム還元剤として
用いているウラナス−ヒドラジンの代りに分離剤として
硝酸−乳酸の混合溶液を用いたものであり、１２段の小
型ミキサ・セトラを使用している。This process uses a mixed solution of nitric acid and lactic acid as a separating agent in place of the uranus-hydrazine used as a plutonium reducing agent in the currently used light water reactor type Burex wet reprocessing process, and is carried out in 12 stages. A small mixer/settler is used.
各供給液の条件について、まとめて簡略化して表わせば
第３図および次表に示す通りである。The conditions for each supply liquid are summarized and simplified as shown in FIG. 3 and the following table.
この条件で連続抽出試験を行った。試験結果を第３図に
示す。A continuous extraction test was conducted under these conditions. The test results are shown in Figure 3.
また、ミキサ・セトラ内のウラン、プルトニウムの濃度
を第４図に示す。Furthermore, the concentrations of uranium and plutonium in the mixer settler are shown in Figure 4.
これらの結果から判るように、ＴＢＰ中に含まれていた
プルトニウムは、乳酸−硝酸混合溶液中に約５倍に濃縮
され９９φ以上の収率で分離回収されていた。As can be seen from these results, the plutonium contained in TBP was concentrated approximately five times in the lactic acid-nitric acid mixed solution and was separated and recovered at a yield of 99φ or more.
一方、プルトニウム中のウランは１５■／ｌでほとんど
がＴＢＰ中に抽出されたまま回収されている。On the other hand, the amount of uranium in plutonium was 15 μ/l, and most of it was recovered as extracted into TBP.
このように、乳酸を従来用いられている洗浄液としての
硝酸に少量添加することのみで、ウラナスを用いること
なく、極めて効率よくウランとプルトニウムの分離を実
現できる。In this way, by simply adding a small amount of lactic acid to the conventionally used nitric acid as a cleaning solution, uranium and plutonium can be separated extremely efficiently without using uranium.
しかも、プルトニウムは約５倍に濃縮されつつ回収され
る結果を得た。Moreover, the plutonium was recovered while being concentrated approximately five times.
なお、プルトニウム製品側の硝酸濃度は１モルで、プル
トニウムの加水分解等の問題も生じないことが判った。The nitric acid concentration on the plutonium product side was 1 mol, and it was found that no problems such as plutonium hydrolysis would occur.
なお、この試験においては、前記したような条件で約１
０時間の連続運転を行い、入口、出口の各濃度の物質収
支がとれたことを確認したのち各段の濃度を測定し、分
離効率０回収効率などを求めた。In addition, in this test, approximately 1
Continuous operation was performed for 0 hours, and after confirming that the mass balance of each concentration at the inlet and outlet was achieved, the concentration at each stage was measured, and the separation efficiency, 0 recovery efficiency, etc. were determined.
また、試験において用いた分離液の濃度は、分離係数を
悪化させない範囲でできるだけ低濃度とした。Furthermore, the concentration of the separation liquid used in the test was kept as low as possible without deteriorating the separation coefficient.
以上本発明の実施例について詳述したが、本発明はこれ
らの実施例に限定されるものでは無滴なく、特許請求の
範囲に記載した範囲内で種々の変更が可能である。Although the embodiments of the present invention have been described in detail above, the present invention is not limited to these embodiments, and various modifications can be made within the scope of the claims.
本発明は上記のように構成されたプルトニウムの分離法
であるから、還元剤の添加や電気的還元操作を伴うこと
なく抽出操作のみによって極めて容易にウランとプルト
ニウムを分離することができ、分離性能もウラナス法と
同じ位に良好であり、分離工程を更に簡略化し得、しか
も、分離したプルトニウムは含まれている硝酸濃度を調
整するのみで容易に抽出され分離剤と分離することがで
きる等数々のすぐれた効果を奏しうるものである。Since the present invention is a plutonium separation method configured as described above, uranium and plutonium can be separated extremely easily by only an extraction operation without adding a reducing agent or an electrical reduction operation, and the separation performance is This method is as good as the Uranas method, and the separation process can be further simplified, and the separated plutonium can be easily extracted and separated from the separating agent by simply adjusting the concentration of nitric acid contained in it. It can produce excellent effects.
第１図はプルトニウムの分離回収フローシート、第２図
はウラン・プルトニウム分離工程図、第３図は乳酸−硝
酸混合溶液におけるウラン・プルトニウム分離工程にお
ける各供給液の条件並びに回収率（実験値）を示す図、
第４図は抽出器内のウランおよびプルトニウムの濃度を
示す図である。Figure 1 is a plutonium separation and recovery flow sheet, Figure 2 is a uranium/plutonium separation process diagram, and Figure 3 is the conditions and recovery rates (experimental values) for each feed liquid in the uranium/plutonium separation process in a lactic acid-nitric acid mixed solution. A diagram showing
FIG. 4 is a diagram showing the concentrations of uranium and plutonium in the extractor.
Claims (1)
ニウムＰ ｕ （Ｍ Ｏｓ ） ４を含む燐の酸素酸エ
ステルを溶媒とする溶液を、乳酸水溶液と低濃度の硝酸
との混合溶液からなる分離剤と接触させることにより、
硝酸ウラニルを含有する燐の酸素酸を溶媒とする溶液と
、プルトニウムを含有する前記分離剤を溶媒とする溶液
とに分離することを特徴とするウランとプルトニウムと
を分離する方法。1. A solution containing uranyl nitrate UO2 (NO3)2 and plutonium nitrate P u (MOs) 4 using a phosphorus oxyester as a solvent is brought into contact with a separating agent consisting of a mixed solution of an aqueous lactic acid solution and a low concentration nitric acid. By this,
A method for separating uranium and plutonium, comprising separating the solution into a solution containing uranyl nitrate using a phosphorus oxyacid as a solvent and a solution containing plutonium using the separating agent as a solvent.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP53041052A JPS5847040B2 (en) | 1978-04-07 | 1978-04-07 | Plutonium separation method |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP53041052A JPS5847040B2 (en) | 1978-04-07 | 1978-04-07 | Plutonium separation method |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS54133293A JPS54133293A (en) | 1979-10-16 |
| JPS5847040B2 true JPS5847040B2 (en) | 1983-10-20 |
Family
ID=12597625
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP53041052A Expired JPS5847040B2 (en) | 1978-04-07 | 1978-04-07 | Plutonium separation method |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS5847040B2 (en) |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5956038A (en) * | 1982-09-20 | 1984-03-31 | Matsushita Electric Ind Co Ltd | Air-conditioner |
| JPS6479532A (en) * | 1987-09-18 | 1989-03-24 | Matsushita Refrigeration | Airflow direction control system for air-conditioning equipment |
-
1978
- 1978-04-07 JP JP53041052A patent/JPS5847040B2/en not_active Expired
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5956038A (en) * | 1982-09-20 | 1984-03-31 | Matsushita Electric Ind Co Ltd | Air-conditioner |
| JPS6479532A (en) * | 1987-09-18 | 1989-03-24 | Matsushita Refrigeration | Airflow direction control system for air-conditioning equipment |
Also Published As
| Publication number | Publication date |
|---|---|
| JPS54133293A (en) | 1979-10-16 |
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