JP2858805B2 - Reprocessing of spent nuclear fuel by low temperature and high load Purex method - Google Patents

Reprocessing of spent nuclear fuel by low temperature and high load Purex method

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Publication number
JP2858805B2
JP2858805B2 JP21888089A JP21888089A JP2858805B2 JP 2858805 B2 JP2858805 B2 JP 2858805B2 JP 21888089 A JP21888089 A JP 21888089A JP 21888089 A JP21888089 A JP 21888089A JP 2858805 B2 JP2858805 B2 JP 2858805B2
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JP
Japan
Prior art keywords
uranium
extraction
plutonium
organic solvent
nitric 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 - Fee Related
Application number
JP21888089A
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Japanese (ja)
Other versions
JPH0382997A (en
Inventor
勝一 館盛
伸夫 久保
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
NIPPON GENSHIRYOKU KENKYUSHO
Original Assignee
NIPPON GENSHIRYOKU KENKYUSHO
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Priority to JP21888089A priority Critical patent/JP2858805B2/en
Priority to FR9008286A priority patent/FR2651364B1/en
Publication of JPH0382997A publication Critical patent/JPH0382997A/en
Application granted granted Critical
Publication of JP2858805B2 publication Critical patent/JP2858805B2/en
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Classifications

    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
    • G21C19/42—Reprocessing of irradiated fuel
    • G21C19/44—Reprocessing of irradiated fuel of irradiated solid fuel
    • G21C19/46—Aqueous processes, e.g. by using organic extraction means, including the regeneration of these means
    • 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

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Inorganic Compounds Of Heavy Metals (AREA)
  • Extraction Or Liquid Replacement (AREA)

Description

【発明の詳細な説明】 （産業上の利用分野） 本発明は、発電用原子炉等から生ずる使用済核燃料中
に含まれるウラン及びプルトニウムといった有用核燃料
物質並びに、安全上重要なネプツニウムを、ピューレッ
クス法により精製分離回収する、いわゆる再処理技術の
高度化に関するものである。DETAILED DESCRIPTION OF THE INVENTION (Industrial application field) The present invention relates to a useful nuclear fuel material such as uranium and plutonium contained in spent nuclear fuel generated from a nuclear reactor for power generation and the like, and neptunium which is important for safety. The present invention relates to the so-called reprocessing technology for purifying, separating and recovering by a method.

（従来の技術） 従来のピューレックス法再処理は、 基本的には常温での溶媒抽出法に基ずいている。但
し、酸化還元反応あるいは、ウランの逆抽出を行う抽出
器では、反応速度を促進させるため、温度を上昇させて
いる。(Prior Art) The conventional Purex reprocessing is basically based on a solvent extraction method at room temperature. However, in an extractor that performs an oxidation-reduction reaction or back extraction of uranium, the temperature is increased in order to accelerate the reaction rate.

ウランとプルトニウムの安定原子価は、夫々Ｕ（V
I）,Pu（IV）であり、両者とも室温では上記抽出剤に良
く抽出される。そのため、ウランとプルトニウムを相互
分離する際には、Pu（IV）を抽出され難いPu（III）に
還元し、Pu（III）を逆抽出することにより行ってい
る。The stable valences of uranium and plutonium are U (V
I) and Pu (IV), both of which are well extracted by the above extractant at room temperature. Therefore, when mutually separating uranium and plutonium, Pu (IV) is reduced to Pu (III) which is hardly extracted, and Pu (III) is back-extracted.

ネプツニウムは、液性により（IV），（Ｖ），（V
I）価と原子価変動をしやすく、原子価調整が困難であ
るため、現在のところ分離回収技術は確定されておら
ず、種々の方法が検討されている。Neptunium is (IV), (V), (V
I) Since the valence and valency are easy to fluctuate and the valence adjustment is difficult, a separation and recovery technique has not been determined at present, and various methods are being studied.

（発明が解決しようとする課題） ピューレックス法再処理では、ウランとプルトニウム
を相互分離するために、後者をIII価に還元する必要が
あり、そのためにFe（III）、Ｕ（IV）あるいは硝酸ヒ
ドロキシルアミン（HAN）といった化学試薬を用いた
り、電気化学的還元法によっている。(Problems to be Solved by the Invention) In the Purex reprocessing, in order to separate uranium and plutonium from each other, it is necessary to reduce the latter to a valence of III, and therefore, Fe (III), U (IV), or nitric acid It uses chemical reagents such as hydroxylamine (HAN) or electrochemical reduction.

これらの還元法は、 複雑な工程制御と装置を必要とするばかりでなく、 Pu（III）をPu（IV）に戻す酸化装置をも要し、 廃棄物量の増大にも帰している また、現在迄知られている、再処理工程におけるネプ
ツニウムの回収法では、最も反応性の高いプルトニウム
とネプツニウムの共存下で、酸化還元反応を利用してウ
ラン、プルトニウム、ネプツニウムの相互分離を試みて
いる。その結果、 これら三つの元素が関与する酸化還元反応は極めて
複雑となり、各元素を収率、純度で分離する事は困
難である。These reduction methods not only require complicated process control and equipment, but also require an oxidizer for returning Pu (III) to Pu (IV), which results in an increase in the amount of waste. In the known method of recovering neptunium in a reprocessing step, uranium, plutonium, and neptunium are mutually separated using a redox reaction in the presence of the most reactive plutonium and neptunium. As a result, the oxidation-reduction reaction involving these three elements becomes extremely complicated, and it is difficult to separate each element in yield and purity.

（課題を解決するための手段） 本発明では、ウランとプルトニウムとを相互分離する
ために、ｉ）抽出系の温度を０〜５℃に低減し、ii）有
機溶媒中のウラン濃度を増大する事により、プルトニウ
ムの抽出分配比を低下させ、かつウランの分配比を増大
させるという効果を利用して、原子価の調整・制御を行
うことなく簡単に両元素を分離出来る。従って、誤操作
等によるトラブルも低減化出来るものである。(Means for Solving the Problems) In the present invention, in order to mutually separate uranium and plutonium, i) reduce the temperature of the extraction system to 0 to 5 ° C, and ii) increase the uranium concentration in the organic solvent. Thus, the two elements can be easily separated without adjusting and controlling the valence by utilizing the effect of decreasing the extraction distribution ratio of plutonium and increasing the distribution ratio of uranium. Therefore, troubles due to erroneous operations and the like can be reduced.

また、本発明では、iii）最初に、プルトニウムを低
温・高負荷工程によって分離回収するので、後に残った
ウランとネプツニウムの分離は、容易に出来る。Further, in the present invention, iii) first, plutonium is separated and recovered by a low-temperature and high-loading process, so that the remaining uranium and neptunium can be easily separated.

すなわち、本発明による方法は、抽出工程の温度低減
化と有機溶媒中のウラン濃度の増大により、使用済核燃
料中のウラン，プルトニウム，ネプツニウムといったア
クチニド元素を、簡便かつ、安全に分離・回収出来るも
のである。That is, the method according to the present invention can easily and safely separate and recover actinide elements such as uranium, plutonium and neptunium in spent nuclear fuel by reducing the temperature of the extraction step and increasing the uranium concentration in the organic solvent. It is.

（実施例） 本発明者らは、リン酸トリブチル−硝酸抽出系におけ
るウラン、プルトニウムの分配比を、種々の条件下で測
定し、さらに、いくつかの抽出サイクル工程を行ったと
ころ、低温・負荷抽出法が、ウラン、プルトニウム、
ネプツニウムの相互分離に極めて有効な事を見出した。(Examples) The present inventors measured the distribution ratio of uranium and plutonium in a tributyl phosphate-nitric acid extraction system under various conditions, and further performed several extraction cycle steps. Extraction method is uranium, plutonium,
It has been found that it is extremely effective for mutual separation of neptunium.

実施例１ ウラン（VI）、プルトニウム（IV）の分配比について
の温度及びウラン濃度の依存性を検討するために、次の
条件下で抽出操作を行った。Example 1 In order to examine the dependence of the distribution ratio of uranium (VI) and plutonium (IV) on the temperature and the uranium concentration, an extraction operation was performed under the following conditions.

30vol％リン酸トリブチル−ドデカンと２−3M硝酸抽
出系におけるＵ（VI）とPu（IV）の分配比を、０、５、
10、15、20、25℃の各種条件下で測定したところ、温度
が低いほど、Pu（IV）の分配比は減少し、Ｕ（VI）の分
配比が増大する事、および、有機溶媒中のウラン濃度が
高いほど、両者の分配比は減少する事がわかった。The partition ratio of U (VI) and Pu (IV) in a 30 vol% tributyl phosphate-dodecane and 2-3M nitric acid extraction system was set to 0, 5,
When measured under various conditions of 10, 15, 20, and 25 ° C, the lower the temperature, the lower the partition ratio of Pu (IV) and the higher the partition ratio of U (VI). It was found that the higher the uranium concentration, the lower the distribution ratio between the two.

そして、この30vol％リン酸トリブチル（TBP）−硝酸
抽出系において、系の温度及び有機溶媒中のウラン濃度
がウラン（VI）、プルトニウム（IV）、の分配比に及ぼ
す影響を示したものが第１図（ａ）−（ｃ）である。In this 30 vol% tributyl phosphate (TBP) -nitric acid extraction system, the effect of the system temperature and the uranium concentration in the organic solvent on the distribution ratio of uranium (VI) and plutonium (IV) was shown. 1 (a)-(c).

実施例２ 主要な核分裂生成物核種の、負荷ウラン条件下にお
ける分配比の温度依存性を検討するために、次の条件下
で抽出操作を行った。Example 2 In order to examine the temperature dependence of the distribution ratio of major fission product nuclides under the conditions of loaded uranium, an extraction operation was performed under the following conditions.

30vol％リン酸トリブチル−ドデカンと２、3M硝酸抽
出系におけるRu,Zr,Nb,Ceの分配比について、それらの
温度依然性を調べた結果、低温・ウラン負荷条件下の
多段抽出工程では、特にZr,Nbで顕著な除染係数の向上
が認められた。As a result of examining the temperature dependence of the distribution ratio of Ru, Zr, Nb, and Ce in a 30 vol% tributyl phosphate-dodecane and 2,3 M nitric acid extraction system, the multi-stage extraction process under low temperature and uranium loading conditions was particularly Zr and Nb significantly improved the decontamination coefficient.

そして、この30vol％TBP−硝酸抽出系における、主要
な核分裂生成物核種の分配比についての、温度、有機溶
媒中のウラン濃度及び硝酸濃度に対する依存性を示した
ものが第２図（ａ），（ｂ）である。FIG. 2 (a) shows the dependence of the distribution ratio of the major fission product nuclides on the temperature, the uranium concentration in the organic solvent and the nitric acid concentration in the 30 vol% TBP-nitric acid extraction system. (B).

実施例３ ウランとプルトニウムとの低温、高負荷工程による相
互分離を第３図（ａ）の抽出工程条件下で行った。Example 3 Uranium and plutonium were separated from each other by a low-temperature, high-load process under the extraction process conditions shown in FIG. 3 (a).

有機溶媒TBP200部が前記工程の第１抽出段に導入さ
れ、1M硝酸溶液50部が第12抽出段に導入され、160g/
のウラン及び2g/のプルトニウムを含有する2M硝酸溶
液100部が第３抽出段に導入され、抽出温度５℃におい
てTBP−硝酸抽出系中で向流抽出処理される。かかる抽
出処理において、80g/のウラン及び３×10-4g/のプ
ルトニウムを含有する有機溶媒抽出液が第12抽出段から
得られ、1.33g/のプルトニウム及び0.13g/のウラン
を含有する硝酸溶液が第１抽出段から得られる。200 parts of organic solvent TBP were introduced into the first extraction stage of the above process, 50 parts of 1M nitric acid solution was introduced into the twelfth extraction stage, and 160 g /
100 parts of a 2M nitric acid solution containing uranium and 2 g / plutonium are introduced into the third extraction stage and subjected to countercurrent extraction in a TBP-nitric acid extraction system at an extraction temperature of 5 ° C. In such an extraction process, an organic solvent extract containing 80 g / uranium and 3 × 10 −4 g / plutonium was obtained from the twelfth extraction stage, and nitric acid containing 1.33 g / plutonium and 0.13 g / uranium. A solution is obtained from the first extraction stage.

その結果得られたウラン、プルトニウム等の抽出器内
濃度分布は第３図（ｂ）に示されるとおりであり、５℃
に維持した抽出工程では、10wt％以下のウランを含むプ
ルトニウム製品と、1.0mg/以下のプルトニウムを含む
ウラン製品が得られる事がわかった。The resulting concentration distribution of uranium, plutonium and the like in the extractor is as shown in FIG.
It was found that, in the extraction process maintained at a temperature of, a plutonium product containing 10 wt% or less of uranium and a uranium product containing 1.0 mg / or less of plutonium were obtained.

実施例４ ウランとネプツニウムの相互分離を第４図（ａ）の抽
出工程条件下で行った。Example 4 Uranium and neptunium were separated from each other under the extraction step conditions shown in FIG. 4 (a).

有機溶媒TBP20部が前記工程の第１抽出段に導入さ
れ、0.5M硝酸溶液20部が第12抽出段に導入され、85g/
のウランと0.1g/のネプツニウムと0.02Mの硝酸とを含
有した。TBP有機溶媒100部が第５抽出段に導入され、そ
して2.5M硝酸溶液50部が第６抽出段に導入され、この中
で向流抽出処理される。かかる抽出処理において、70.8
g/のウラン（VI）、8.5×10-5g/及び0.02Mの硝酸を
含有する有機溶媒抽出液が第12抽出段から得られ、0.25
g/のネプツニウム（IV）、0.0075g/のウラン及び1.
5Mの硝酸を含有する硝酸抽出液が第１抽出段から得られ
る。20 parts of organic solvent TBP were introduced into the first extraction stage of the above process, 20 parts of 0.5M nitric acid solution were introduced into the twelfth extraction stage, and 85 g /
Uranium, 0.1 g / neptunium and 0.02 M nitric acid. 100 parts of TBP organic solvent are introduced into the fifth extraction stage, and 50 parts of a 2.5M nitric acid solution are introduced into the sixth extraction stage, where they are subjected to countercurrent extraction. In this extraction process, 70.8
g / uranium (VI), 8.5 × 10 −5 g / and an organic solvent extract containing 0.02 M nitric acid were obtained from the twelfth extraction stage,
g / neptunium (IV), 0.0075 g / uranium and 1.
A nitric acid extract containing 5M nitric acid is obtained from the first extraction stage.

その結果得られたウラン、ネプツニウム等の抽出器内
濃度分布は第４図（ｂ）に示されるとおりであり、99.9
％以上のネプツニウムが5wt％以下のウランを含んで分
離回収される事がわかった。The resulting concentration distribution of uranium, neptunium, etc. in the extractor is as shown in FIG.
It was found that more than 5% by weight of neptunium was separated and recovered, including less than 5% by weight of uranium.

（発明の効果） 再処理施設において、使用済核燃料の溶解液を、ピュ
ーレックス法によって処理する際にもたらされるもので
あり、それは次のとおりである。(Effect of the Invention) This is brought about when a solution of spent nuclear fuel is treated by the Purex method in a reprocessing facility, and is as follows.

溶解液中に含まれる有用な核燃料物質であるウラン
とプルトニウムを、夫々、収率・純度で分離・回収
することができる。Uranium and plutonium, which are useful nuclear fuel substances contained in the solution, can be separated and recovered with a yield and purity, respectively.

ウランとプルトニウムの相互分離は、単に工程の温
度を低下させ、有機溶媒中のウラン濃度を増大させる事
により、達成出来る。Mutual separation of uranium and plutonium can be achieved by simply lowering the temperature of the process and increasing the uranium concentration in the organic solvent.

長期にわたる毒性から、ウラン及び廃棄物中に含ま
れてはならないネプツニウムを収率で分離・回収出来
る。Due to long-term toxicity, neptunium which must not be contained in uranium and waste can be separated and recovered in a yield.

【図面の簡単な説明】[Brief description of the drawings]

第１図（ａ），（ｂ），（ｃ）は30vol％リン酸トリブ
チル（TBP）−硝酸抽出系において、系の温度及び有機
溶媒中のウラン濃度が、ウラン（VI）とプルトニウム
（IV）の分配比に及ぼす影響を示すグラフである。 第２図（ａ），（ｂ）は30vol％TBP−硝酸抽出系におけ
る、主要な核分裂生成物核種の分配比の、温度、有機溶
媒中ウラン濃度、及び硝酸濃度依存性をグラフである。 第３図（ａ）は30vol％TBP−硝酸抽出系の、低温（５
℃）・負荷Ｕ（VI）−Pu（IV）分離工程フローシート
であり、第３図（ｂ）はウラン、プルトニウム等の濃度
分布を示すグラフである。 第４図（ａ）は30vol％TBP−硝酸抽出系の、ウランとネ
プツニウムの分離工程フローシートであり、第４図
（ｂ）はウラン、ネプツニウム等の濃度分布を示すグラ
フである。FIGS. 1 (a), (b) and (c) show that in a 30 vol% tributyl phosphate (TBP) -nitric acid extraction system, the temperature of the system and the uranium concentration in the organic solvent are uranium (VI) and plutonium (IV). 4 is a graph showing the effect on the distribution ratio. FIGS. 2 (a) and 2 (b) are graphs showing the dependence of the distribution ratio of the main fission product nuclides on the temperature, the uranium concentration in the organic solvent, and the nitric acid concentration in the 30 vol% TBP-nitric acid extraction system. FIG. 3 (a) shows the low temperature (5 vol.
C) / load U (VI) -Pu (IV) separation step flow sheet, and FIG. 3 (b) is a graph showing the concentration distribution of uranium, plutonium and the like. FIG. 4 (a) is a flow sheet of a 30 vol% TBP-nitric acid extraction system for separating uranium and neptunium, and FIG. 4 (b) is a graph showing the concentration distribution of uranium, neptunium and the like.

フロントページの続き (58)調査した分野(Int.Cl.6，ＤＢ名) G21C 19/46 ＪＩＣＳＴ（ＪＯＩＳ)Continuation of the front page (58) Field surveyed (Int.Cl. 6 , DB name) G21C 19/46 JICST (JOIS)

Claims (1)

(57)【特許請求の範囲】(57) [Claims] 【請求項１】使用済み核燃料廃液中に含まれるウラン、
プルトニウム、ネブツニウムを溶媒抽出法により分離回
収する方法において、 （ａ） ウランとプルトニウムとを相互分離するため
に、抽出系の温度を０〜５℃に低減して有機溶媒中のウ
ランの抽出分配比を増大させるとともにプルトニウムの
抽出分配比を低下させることにより、その抽出系の一端
からウラン抽出有機溶媒を供給し、その他端から硝酸溶
液を供給し、抽出系の有機溶媒の供給側に近い位置に前
記核燃料廃液を供給し、プルトニウムを硝酸溶液に抽出
してプルトニウム含有溶液を前記一端側から取り出し、
ウラン及びネブツニウムを有機溶媒に抽出してウラン及
びネブツニウム含有溶液を前記他端側から取り出し、次
に （ｂ） ウランとネブツニウムとを相互分離するため
に、他の抽出系の一端からウラン抽出有機溶媒を供給
し、その他端から硝酸溶液を供給し、抽出系の中央部に
前記ウラン及びネブツニウム含有溶液を供給し、ネブツ
ニウム含有溶液を前記一端側から取り出し、ウラン含有
精製溶液を前記他端側から取り出す方法。1. Uranium contained in spent nuclear fuel waste liquid,
In a method for separating and recovering plutonium and nebutunium by a solvent extraction method, (a) in order to mutually separate uranium and plutonium, the extraction system temperature is reduced to 0 to 5 ° C. to extract and distribute uranium in an organic solvent; By increasing the extraction distribution ratio of plutonium and supplying the uranium extraction organic solvent from one end of the extraction system, and supplying the nitric acid solution from the other end of the extraction system, at a position close to the supply side of the extraction system organic solvent. Supplying the nuclear fuel waste liquid, extracting plutonium into nitric acid solution and removing the plutonium-containing solution from the one end side,
Uranium and nebutunium are extracted into an organic solvent to remove the uranium and nebutunium-containing solution from the other end, and then (b) a uranium extraction organic solvent is extracted from one end of another extraction system to mutually separate uranium and nebutunium. And the nitric acid solution is supplied from the other end, the uranium and nebutunium-containing solution is supplied to the center of the extraction system, the nebutunium-containing solution is taken out from the one end, and the uranium-containing purified solution is taken out from the other end. Method.
JP21888089A 1989-08-25 1989-08-25 Reprocessing of spent nuclear fuel by low temperature and high load Purex method Expired - Fee Related JP2858805B2 (en)

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FR9008286A FR2651364B1 (en) 1989-08-25 1990-06-29 PROCESS FOR THE RE-TREATMENT OF USED NUCLEAR FUELS.

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LU84016A1 (en) * 1982-03-15 1982-07-08 Euratom METHOD FOR THE RECOVERY OF PLUTONIUM FROM NITROUS ACID SOLUTIONS
US4528165A (en) * 1984-06-13 1985-07-09 The United States Of America As Represented By The United States Department Of Energy Separation of uranium from technetium in recovery of spent nuclear fuel
EP0251399A1 (en) * 1986-06-23 1988-01-07 "Centre d'Etude de l'Energie Nucléaire", "C.E.N." Process for separating or recovering plutonium, and plutonium obtained thereby

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