JP5133140B2 - Spent fuel treatment method - Google Patents

Spent fuel treatment method Download PDF

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JP5133140B2
JP5133140B2 JP2008149552A JP2008149552A JP5133140B2 JP 5133140 B2 JP5133140 B2 JP 5133140B2 JP 2008149552 A JP2008149552 A JP 2008149552A JP 2008149552 A JP2008149552 A JP 2008149552A JP 5133140 B2 JP5133140 B2 JP 5133140B2
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uranium
spent fuel
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崇 阿部
朗 笹平
国義 星野
哲生 深澤
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Hitachi GE Nuclear Energy Ltd
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    • 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
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    • 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
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    • Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
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Description

本発明は熱中性子炉から排出される使用済燃料に含まれるウランを溶媒抽出法を用いて取り出す使用済燃料の処理方法に関する。   The present invention relates to a spent fuel processing method for extracting uranium contained in spent fuel discharged from a thermal neutron reactor using a solvent extraction method.

従来の燃料サイクルシステムでは、軽水炉用再処理施設および高速増殖炉用再処理施設を用いて、軽水炉使用済燃料処理、高速増殖炉用燃料製造および高速増殖炉用使用済燃料処理を行う仕組みとなっている。このシステムにおいて、軽水炉用再処理施設は、軽水炉発電所から発生する軽水炉使用済燃料を再処理し、回収ウランと高速増殖炉用燃料製造を行っている。また、高速増殖炉用再処理施設は、高速増殖炉使用済燃料の再処理と高速増殖炉用燃料の製造を行っている（非特許文献１等参照）。   The conventional fuel cycle system uses a light water reactor reprocessing facility and a fast breeder reactor reprocessing facility to perform light water reactor spent fuel processing, fast breeder reactor fuel manufacturing, and fast breeder reactor spent fuel processing. ing. In this system, the light water reactor reprocessing facility reprocesses the light water reactor spent fuel generated from the light water reactor power plant to produce recovered uranium and fuel for the fast breeder reactor. In addition, the fast breeder reactor reprocessing facility reprocesses the fast breeder spent fuel and manufactures the fast breeder reactor fuel (see Non-Patent Document 1, etc.).

ところで、熱中性子炉（例えば、軽水炉）から高速中性子炉（例えば、高速増殖炉）への原子力発電の移行時期には、熱中性子炉使用済燃料の再処理と、高速中性子炉使用済燃料の再処理の両方を行う必要がある。しかし、熱中性子炉の再処理施設でウラン、ウランとプルトニウムの混合物の抽出と燃料への加工を行いながら、高速中性子炉の再処理施設でもウラン、ウランとプルトニウムの混合物の抽出と燃料への加工を行うと、どちらの施設も高機能化する必要があり施設が大型化し、高コスト化してしまう。この種の課題の解決を図った技術としては、熱中性子炉使用済燃料からウランを除去して容積及び重量を減少させる減容処理施設を設け、ここで減容した使用済燃料を高速中性子炉使用済燃料の再処理施設に供給することで、プルトニウムとウランの抽出及び加工を図ったものがある（特許文献１等参照）。   By the way, during the transition of nuclear power generation from a thermal neutron reactor (for example, a light water reactor) to a fast neutron reactor (for example, a fast breeder reactor), reprocessing of the spent fuel for the thermal neutron reactor and reprocessing of the spent fuel for the fast neutron reactor Both processes need to be done. However, while extracting uranium, a mixture of uranium and plutonium and processing it into fuel at the reprocessing facility of the thermal neutron reactor, extracting the uranium, uranium and plutonium mixture and processing it into the fuel at the reprocessing facility of the fast neutron reactor If this is done, both facilities will need to be highly functional, making the facilities larger and costly. As a technology for solving this type of problem, a volume reduction treatment facility that reduces the volume and weight by removing uranium from the thermal neutron reactor spent fuel is provided, and the spent fuel reduced here is used as a fast neutron reactor. There is one in which plutonium and uranium are extracted and processed by supplying them to a spent fuel reprocessing facility (see Patent Document 1, etc.).

この技術のように、使用済燃料からウランを除去し、容積、重量を減少させるウラン粗分離法（Ｕ粗分離法）の具体例としては、（１）晶析法、（２）沈殿法、（３）ＴＢＰ抽出法（ＰＵＲＥＸ法の条件調整）、及び（４）モノアミド抽出法の４種類の方法がある（非特許文献２等参照）。このうちＴＢＰ抽出法は、現在実施されているＰＵＲＥＸ法再処理の条件を変更することで粗分離を達成できるので、他の３種の方法に比べて機器開発及びプラント建設の期間が短く、短期間に実用化が望める方法である。このＴＢＰ抽出法を利用した技術としては、使用済燃料を溶解した硝酸溶液を還元条件下でＴＢＰ（トリブチル燐酸）と接触させ、そのＴＢＰ中にウランを抽出する方法がある（特許文献２等参照）。   Specific examples of the uranium rough separation method (U rough separation method) for removing uranium from spent fuel and reducing the volume and weight as in this technique include (1) crystallization method, (2) precipitation method, There are four types of methods: (3) TBP extraction method (condition adjustment of PUREX method) and (4) monoamide extraction method (see Non-Patent Document 2, etc.). Of these, the TBP extraction method can achieve rough separation by changing the conditions of the currently implemented PUREX method reprocessing, so the period of equipment development and plant construction is shorter and shorter than the other three methods. It is a method that can be practically used in the meantime. As a technique using this TBP extraction method, there is a method in which a nitric acid solution in which spent fuel is dissolved is brought into contact with TBP (tributyl phosphoric acid) under reducing conditions, and uranium is extracted from the TBP (see Patent Document 2, etc.). ).

「原子力のすべて」編集委員会著，「原子力のすべて」，国立印刷局"All about nuclear power" Editorial Board, "All about nuclear power", National Printing Bureau （社）日本原子力学会 再処理・リサイクル部会，「第５回再処理リサイクルセミナーテキスト」，ｐ９７Japan Atomic Energy Society Reprocessing / Recycling Subcommittee, “5th Reprocessing Recycling Seminar Text”, p.97 特開２００６−２７５６３８号公報JP 2006-275638 A 特開平７−３５８９４号公報Japanese Patent Laid-Open No. 7-35894

しかしながら、特許文献２に記載されている上記方法では、抽出を調整するために還元剤や新抽出剤などの試薬を添加して還元条件下におく必要がある。そのため、添加した試薬の後続処理への影響が未知であり、現行プラントでの使用にはリスクを伴う可能性があることを完全に否定できない。   However, in the above-described method described in Patent Document 2, it is necessary to add a reagent such as a reducing agent or a new extracting agent under reducing conditions in order to adjust the extraction. For this reason, the influence of the added reagent on the subsequent processing is unknown, and it cannot be completely denied that there is a risk of using it in the current plant.

本発明の目的は、溶媒抽出法において、還元剤を添加することなく使用済燃料に含まれるウランを取り出すことができる使用済燃料の処理方法を提供することにある。   An object of the present invention is to provide a spent fuel processing method capable of extracting uranium contained in spent fuel without adding a reducing agent in a solvent extraction method.

（１）本発明は、上記目的を達成するために、熱中性子炉から排出される使用済燃料に含まれるウランを溶媒抽出法を用いて取り出す使用済燃料の処理方法において、使用済燃料を硝酸に溶解させた溶解液に対して、当該溶解液中のウランとプルトニウムの全量を抽出するために最低限必要な流量より少ないＴＢＰを含む有機溶媒を接触させて、当該有機溶媒中にウランを抽出し、プルトニウムの大部分とウランの一部を前記溶解液中に残すことで前記使用済燃料に含まれるウランの大部分を取り出すものとする。   (1) In order to achieve the above object, the present invention provides a spent fuel treatment method in which uranium contained in spent fuel discharged from a thermal neutron reactor is extracted using a solvent extraction method. Extract the uranium into the organic solvent by contacting the organic solvent containing less TBP than the minimum required flow rate to extract the total amount of uranium and plutonium in the dissolved solution. Then, most of the uranium contained in the spent fuel is taken out by leaving most of the plutonium and a part of uranium in the solution.

（２）上記（１）は、好ましくは、前記ウランを抽出した有機溶媒に対して、前記溶解液以上のウラン濃度を有する洗浄用硝酸溶液を接触させて、前記ウランを抽出した有機溶媒に混入したプルトニウムを前記洗浄用硝酸溶液に回収し、前記有機溶媒中のウランを浄化するものとする。   (2) The above (1) is preferably mixed with the organic solvent from which the uranium has been extracted by contacting the organic solvent from which the uranium has been extracted with a cleaning nitric acid solution having a uranium concentration equal to or higher than that of the solution. The plutonium thus collected is recovered in the cleaning nitric acid solution to purify uranium in the organic solvent.

本発明によれば、還元剤を添加することなく使用済燃料に含まれるウランの大部分を取り出すことができる。   According to the present invention, most of the uranium contained in the spent fuel can be taken out without adding a reducing agent.

以下、本発明の実施の形態を図面を用いて説明する。   Hereinafter, embodiments of the present invention will be described with reference to the drawings.

図１は本発明の実施の形態である使用済燃料の処理工程のフロー図である。   FIG. 1 is a flowchart of spent fuel processing steps according to an embodiment of the present invention.

この図に示す処理工程はＰＵＲＥＸ法を利用したものである。ＰＵＲＥＸ法は、使用済燃料を溶解した溶解液（硝酸溶液）にＴＢＰを含む有機溶媒を接触させ、ウランとプルトニウムを有機溶媒に抽出する方法である。ＰＵＲＥＸ法では有機溶媒を他の硝酸で洗浄し、残余の不純物を当該他の硝酸に取り込んでウランとプルトニウムを洗浄する。   The processing steps shown in this figure utilize the PUREX method. The PUREX method is a method in which an organic solvent containing TBP is brought into contact with a solution (nitric acid solution) in which spent fuel is dissolved, and uranium and plutonium are extracted into the organic solvent. In the PUREX method, the organic solvent is washed with other nitric acid, and the remaining impurities are taken into the other nitric acid to wash uranium and plutonium.

図１に示した処理工程は、熱中性子炉から排出された使用済燃料（使用済原子燃料）５からウランを粗分離するために、溶解槽１と、粗分離抽出器２と、洗浄器３と、逆抽出用抽出器４を利用している。   The processing step shown in FIG. 1 includes a dissolution tank 1, a coarse separation extractor 2, and a washing device 3 in order to roughly separate uranium from spent fuel (spent nuclear fuel) 5 discharged from a thermal neutron reactor. And the extractor 4 for back extraction is utilized.

溶解槽１は、使用済燃料５を溶解用硝酸６に溶解させて溶解液７を得るためのものである。溶解槽１で得られる溶解液７には、ウラン（Ｕ）と、プルトニウム（Ｐｕ）と、不純物が含まれている。なお、ここでいう不純物とは、マイナーアクチナイドと核分裂生成物である。   The dissolution tank 1 is for obtaining a solution 7 by dissolving the spent fuel 5 in the nitric acid 6 for dissolution. The solution 7 obtained in the dissolution tank 1 contains uranium (U), plutonium (Pu), and impurities. The impurities here are minor actinides and fission products.

粗分離抽出器２は、溶解槽１からの溶解液７に対してトリブチル燐酸（ＴＢＰ）を含む有機溶媒８を接触させて、有機溶媒中にウランを抽出するものである。ここでウランを抽出した有機溶媒（抽出後有機溶媒）９は、洗浄器３に導入される。   The coarse separation extractor 2 is an apparatus in which an organic solvent 8 containing tributyl phosphoric acid (TBP) is brought into contact with the solution 7 from the dissolution tank 1 to extract uranium in the organic solvent. Here, the organic solvent 9 (the organic solvent after extraction) from which uranium has been extracted is introduced into the cleaning device 3.

洗浄器３は、粗分離抽出器２からの抽出後有機溶媒９に対して洗浄用硝酸溶液（洗浄液）１０を接触させて、抽出後有機溶媒９に混入したプルトニウムを回収するものである。ここで抽出後有機溶媒９から洗浄液１０に放出されたプルトニウムは、その他の不純物とともに回収洗浄液１５に含まれ、粗分離抽出器２に戻される。粗分離抽出器２に戻されたプルトニウムは、最終的に抽出後溶解液１３に含まれ、減容した使用済燃料１６となって再生処理施設等（図示せず）に供給される。   The washer 3 is for bringing a washing nitric acid solution (washing solution) 10 into contact with the organic solvent 9 after extraction from the coarse separation extractor 2 to recover plutonium mixed in the organic solvent 9 after extraction. Here, the plutonium released from the organic solvent 9 to the cleaning liquid 10 after extraction is contained in the recovered cleaning liquid 15 together with other impurities and returned to the coarse separation extractor 2. The plutonium returned to the coarse separation extractor 2 is finally contained in the solution 13 after extraction, and is used as a reduced spent fuel 16 to be supplied to a regeneration facility (not shown).

逆抽出用抽出器４は、洗浄器３で浄化された洗浄後有機溶媒１１に対して、希薄な硝酸などの逆抽出液を接触させて、洗浄後有機溶媒１１に含まれたウランとプルトニウムを逆抽出液に放出するためのものである。逆抽出液に放出されたウランとプルトニウムは、回収ウラン（粗分離ウラン）１４として取り出される。これにより使用済燃料５からのウランの回収が完了する。   The extractor 4 for back extraction is made to contact a back extract solution such as dilute nitric acid with the washed organic solvent 11 purified by the washing device 3 to remove uranium and plutonium contained in the washed organic solvent 11. For release into the back extract. Uranium and plutonium released into the back extract are taken out as recovered uranium (crude separated uranium) 14. Thereby, the recovery of uranium from the spent fuel 5 is completed.

上記のように構成される本実施の形態の処理工程は、粗分離抽出器２および洗浄器３における条件を下記のように調節することにより、抽出後溶解液１３にプルトニウムを残すとともに、回収ウラン１４のＰｕ含有量を低減させている。以下、これについて説明する。   The processing step of the present embodiment configured as described above is to adjust the conditions in the coarse separation extractor 2 and the washer 3 as follows, thereby leaving plutonium in the solution 13 after extraction, and recovering uranium. 14 Pu content is reduced. This will be described below.

本実施の形態において、粗分離抽出器２に導入される有機溶媒８には、溶解液７中のウラン及びプルトニウムの全量を抽出するために最低限必要な流量より少ないＴＢＰが含まれている。このように有機溶媒８中のＴＢＰの流量を調節すると、図２に示すように、プルトニウムの大部分とウランの一部を溶解液中に残すことができる。   In the present embodiment, the organic solvent 8 introduced into the coarse separation extractor 2 contains TBP less than the minimum flow rate necessary for extracting the total amount of uranium and plutonium in the solution 7. When the flow rate of TBP in the organic solvent 8 is adjusted as described above, most of plutonium and a part of uranium can be left in the solution as shown in FIG.

図２は粗分離抽出器２におけるウラン回収率とプルトニウムのウランへの同伴率の関係を示す図である。この図は発明者らが得た知見に基づいて作成されたものである。   FIG. 2 is a graph showing the relationship between the uranium recovery rate in the coarse separation extractor 2 and the rate of plutonium entrainment with uranium. This figure was created based on the knowledge obtained by the inventors.

上記のように、粗分離抽出器２に導入された溶解液７は、有機溶媒８と接触して有機溶媒８中にウランとプルトニウムを放出する。このとき、有機溶媒８中に含まれるＴＢＰの流量が充分にあれば、ウランもプルトニウムも完全に有機溶媒に移動する。しかし、ＴＢＰの流量がウランの流量より少ないと、ウランはＴＢＰの流量分以上は有機溶媒に移動することができないので、抽出後有機溶媒９のウラン回収率（Ｕ回収率）は低下する。このとき、プルトニウムもＴＢＰの流量が少なければ有機溶媒への移動率は低下するが、その割合はウランより遙かに小さい。この関係を示したのが図２であり、図２はＴＢＰの流量を変えてＵ回収率を変化させたときのプルトニウムのウランへの同伴率を示している。   As described above, the solution 7 introduced into the coarse separation extractor 2 comes into contact with the organic solvent 8 and releases uranium and plutonium into the organic solvent 8. At this time, if the flow rate of TBP contained in the organic solvent 8 is sufficient, uranium and plutonium are completely transferred to the organic solvent. However, if the TBP flow rate is less than the uranium flow rate, uranium cannot move to the organic solvent beyond the TBP flow rate, so the uranium recovery rate (U recovery rate) of the organic solvent 9 after extraction decreases. At this time, if the flow rate of TBP is small, the transfer rate to the organic solvent also decreases, but the ratio is much smaller than uranium. FIG. 2 shows this relationship, and FIG. 2 shows the entrainment ratio of plutonium with uranium when the UBP flow rate is changed and the U recovery rate is changed.

図２において、Ｕ回収率が１の場合はＰｕの回収率も１であり、このときのＴＢＰの流量が溶解液７中のウラン及びプルトニウムの全量を抽出するために最低限必要な流量となる。しかし、ＴＢＰの流量をこの値より少なくしてＵ回収率を１未満にすると、Ｐｕ同伴率はＵ回収率と比較して極めて小さくなる。例えば、Ｕ回収率が０．９のときは、Ｐｕ同伴率は約０．２５となる。これは、ＴＢＰ抽出によるウランの粗分離率が９０％のとき、当該粗分離したウランには、使用済燃料５に含まれていたプルトニウムの２５％が含まれることを示す。粗分離するウランにどの程度のプルトニウムを同伴させるかは、粗分離するウランの用途に応じて決定すれば良い。すなわち、有機溶媒８中に含有させるＴＢＰの流量は、ウランの用途に応じて調節すれば良い。このように、本実施の形態によれば、ＴＢＰの流量を調節して使用済燃料５に含まれたプルトニウムの大部分とウランの一部を溶解液中に残すことで、使用済燃料５に含まれるウランの大部分を取り出せることができる。このようにウランを取り出すと、還元剤等の試薬を添加する必要がないので、後続処理への影響を懸念する必要がなくなり、現行プラントで使用してもその信頼性を維持することができる。   In FIG. 2, when the U recovery rate is 1, the Pu recovery rate is 1, and the TBP flow rate at this time is the minimum flow rate required to extract the total amount of uranium and plutonium in the solution 7. . However, if the TBP flow rate is less than this value and the U recovery rate is less than 1, the Pu entrainment rate is extremely small compared to the U recovery rate. For example, when the U recovery rate is 0.9, the Pu entrainment rate is about 0.25. This indicates that when the crude uranium separation rate by TBP extraction is 90%, the roughly separated uranium contains 25% of the plutonium contained in the spent fuel 5. How much plutonium is to be accompanied with the uranium to be roughly separated may be determined according to the use of the uranium to be roughly separated. That is, what is necessary is just to adjust the flow volume of TBP contained in the organic solvent 8 according to the use of uranium. Thus, according to the present embodiment, by adjusting the TBP flow rate and leaving most of the plutonium contained in the spent fuel 5 and a part of uranium in the solution, Most of the uranium contained can be extracted. When uranium is taken out in this way, it is not necessary to add a reagent such as a reducing agent, so there is no need to worry about the influence on the subsequent processing, and the reliability can be maintained even when used in the current plant.

また、本実施の形態は、上記構成に加え、洗浄後有機溶媒１１に含まれるプルトニウムの量を下記構成により更に低減させている。   In addition to the above configuration, the present embodiment further reduces the amount of plutonium contained in the organic solvent 11 after cleaning by the following configuration.

本実施の形態において、洗浄器３に導入される洗浄液１０にはウランが含有されており、洗浄液１０のウラン濃度は溶解液７のウラン濃度以上に設定されている。この濃度のウランは、粗分離抽出器２におけるウランと同程度かそれ以上に有機溶媒に移動しやすいため、抽出後有機溶媒９に含まれるプルトニウムと置き換わり、そのプルトニウムを洗浄液１０に放出させる効果がある。なお、洗浄器３におけるプルトニウムの放出効果は、洗浄液１０のウラン濃度が高いか、洗浄液の流量が大きいほど高くなる。   In the present embodiment, the cleaning liquid 10 introduced into the cleaning device 3 contains uranium, and the uranium concentration of the cleaning liquid 10 is set to be equal to or higher than the uranium concentration of the solution 7. This concentration of uranium is likely to move to an organic solvent to the same extent or more than uranium in the coarse separation extractor 2, so that it replaces the plutonium contained in the organic solvent 9 after extraction and has the effect of releasing the plutonium into the cleaning liquid 10 is there. Note that the effect of releasing plutonium in the cleaning device 3 increases as the uranium concentration in the cleaning liquid 10 increases or the flow rate of the cleaning liquid increases.

図３は、洗浄器３における洗浄液１０と有機相の流量比と、抽出後有機溶媒９からのプルトニウムの回収率の関係を示す図である。この図も図２同様に発明者らが得た知見に基づいて作成されたものである。なお、ここで洗浄液１０と有機相の流量比とは、洗浄液１０と抽出後有機溶媒９の流量の和に対する洗浄液１０の流量の割合である。   FIG. 3 is a diagram showing the relationship between the flow rate ratio between the cleaning liquid 10 and the organic phase in the cleaning device 3 and the recovery rate of plutonium from the organic solvent 9 after extraction. This figure is also created based on the knowledge obtained by the inventors as in FIG. Here, the flow rate ratio between the cleaning liquid 10 and the organic phase is the ratio of the flow rate of the cleaning liquid 10 to the sum of the flow rates of the cleaning liquid 10 and the organic solvent 9 after extraction.

図３において、流量比を０．２２とすると、抽出後有機溶媒９からのプルトニウムの回収率は９５％となる。このとき、例えば、最初の粗分離抽出器２におけるプルトニウム同伴率が２０〜３０％であった場合には、洗浄器３を通過して、最終的に逆抽出容抽出器４で得られる回収ウラン（粗分離ウラン）１４に含まれるプルトニウムの量は１〜１．５％となる。したがって、この場合には、プルトニウム同伴率が５％以下の粗分離ウラン１４を得ることができる。また、洗浄器３において回収されたプルトニウムと不純物は回収洗浄液１５に含まれて粗分離抽出器２に戻り、最終的に抽出後溶解液１３に含まれる。これにより、全体として使用済燃料から５％以下のＰｕを含むウランを粗分離したことになる。   In FIG. 3, when the flow rate ratio is 0.22, the recovery rate of plutonium from the organic solvent 9 after extraction is 95%. At this time, for example, when the plutonium entrainment ratio in the first coarse separation extractor 2 is 20 to 30%, the recovered uranium finally passes through the washer 3 and is finally obtained by the back extraction extractor 4. The amount of plutonium contained in (roughly separated uranium) 14 is 1 to 1.5%. Therefore, in this case, crude separation uranium 14 having a plutonium entrainment ratio of 5% or less can be obtained. Further, the plutonium and impurities recovered in the cleaning device 3 are contained in the recovered cleaning solution 15 and returned to the coarse separation extractor 2, and finally included in the solution 13 after extraction. As a result, uranium containing 5% or less of Pu is roughly separated from spent fuel as a whole.

このように本実施の形態によれば、洗浄器３においても還元剤等を用いることなくプルトニウムを回収することができるので、回収ウラン１４中のプルトニウム量を更に低減することができる。また、粗分離抽出器２と洗浄器３の条件を変更することで、回収ウラン１４中に含まれるプルトニウム量をウランの用途に合わせて柔軟に調節することもできる。   As described above, according to the present embodiment, since the plutonium can be recovered without using a reducing agent or the like in the washer 3, the amount of plutonium in the recovered uranium 14 can be further reduced. Moreover, the amount of plutonium contained in the recovered uranium 14 can be flexibly adjusted according to the use of uranium by changing the conditions of the coarse separation extractor 2 and the washer 3.

以上説明したように、本実施の形態は、上記の一連の処理により、使用済燃料５を溶解した溶解液７に含まれるウランの大部分を洗浄された回収ウラン１４とする一方で、溶解液７に含まれるプルトニウムと残余のウランを不純物とともに溶解液１３として最初の粗分離抽出工程２から払い出すことができる。   As described above, in the present embodiment, a large part of uranium contained in the solution 7 in which the spent fuel 5 is dissolved is used as the recovered recovered uranium 14 by the above-described series of processes. 7 and the remaining uranium can be discharged from the first coarse separation and extraction step 2 as a solution 13 together with impurities.

なお、上記の説明では、粗分離抽出器２および洗浄器３の両方でプルトニウムを回収する構成について説明したが、ウランの用途によって両方でプルトニウムを回収する必要が無い場合には、いずれか一方でプルトニウムを回収すれば良い。   In the above description, the configuration in which plutonium is recovered by both the coarse separation extractor 2 and the cleaning device 3 has been described. However, if it is not necessary to recover plutonium by both uranium uses, What is necessary is just to collect plutonium.

本発明の実施の形態である使用済燃料の処理工程のフロー図。The flowchart of the processing process of the spent fuel which is embodiment of this invention. 粗分離抽出器におけるウラン回収率とプルトニウムのウランへの同伴率の関係を示す図。The figure which shows the relationship between the uranium recovery rate in a rough separation extractor, and the entrainment rate to the uranium of plutonium. 洗浄器における洗浄液と有機相の流量比と抽出後有機溶媒からのプルトニウムの回収率の関係を示す図The figure which shows the relationship between the flow ratio of the washing liquid and the organic phase in the washer and the recovery rate of plutonium from the organic solvent after extraction

符号の説明Explanation of symbols

２ 粗分離抽出器
３ 洗浄器
４ 逆抽出用抽出器
５ 使用済燃料
６ 溶解用硝酸
７ 溶解液
８ 有機溶媒
９ 抽出後有機溶媒
１０ 洗浄液
１１ 洗浄後有機溶媒
１３ 抽出後溶解液
１５ 回収洗浄液 DESCRIPTION OF SYMBOLS 2 Coarse separation extractor 3 Washer 4 Extractor for back extraction 5 Spent fuel 6 Nitric acid for dissolution 7 Solvent 8 Organic solvent 9 Organic solvent after extraction 10 Washing liquid 11 Organic solvent after washing 13 Solute after extraction 15 Recovery washing liquid

Claims (2)

熱中性子炉から排出される使用済燃料に含まれるウランを溶媒抽出法を用いて取り出す使用済燃料の処理方法において、
使用済燃料を硝酸に溶解させた溶解液に対して、当該溶解液中のウランとプルトニウムの全量を抽出するために最低限必要な流量より少ないＴＢＰを含む有機溶媒を接触させて、当該有機溶媒中にウランを抽出し、
プルトニウムの大部分とウランの一部を前記溶解液中に残すことで前記使用済燃料に含まれるウランの大部分を取り出すことを特徴とする使用済燃料の処理方法。 In the spent fuel processing method for extracting uranium contained in the spent fuel discharged from the thermal neutron reactor using a solvent extraction method,
An organic solvent containing TBP less than the minimum flow rate required to extract the total amount of uranium and plutonium in the dissolved solution is brought into contact with a dissolved solution obtained by dissolving spent fuel in nitric acid, and the organic solvent Extract uranium inside,
A method for treating spent fuel, wherein most of plutonium and a part of uranium are left in the solution to take out most of uranium contained in the spent fuel. 請求項１記載の使用済燃料の処理方法において、
前記ウランを抽出した有機溶媒に対して、前記溶解液以上のウラン濃度を有する洗浄用硝酸溶液を接触させて、前記ウランを抽出した有機溶媒に混入したプルトニウムを前記洗浄用硝酸溶液に回収し、
前記有機溶媒中のウランを浄化することを特徴とする使用済燃料の処理方法。 The spent fuel processing method according to claim 1,
The organic solvent from which the uranium has been extracted is brought into contact with a cleaning nitric acid solution having a uranium concentration equal to or higher than the solution, and the plutonium mixed in the organic solvent from which the uranium has been extracted is recovered in the cleaning nitric acid solution,
A method for treating spent fuel, comprising purifying uranium in the organic solvent.
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