JPH0382997A - Spent nuclear fuel reprocessing method by low temperature and high load purex method - Google Patents
Spent nuclear fuel reprocessing method by low temperature and high load purex methodInfo
- Publication number
- JPH0382997A JPH0382997A JP1218880A JP21888089A JPH0382997A JP H0382997 A JPH0382997 A JP H0382997A JP 1218880 A JP1218880 A JP 1218880A JP 21888089 A JP21888089 A JP 21888089A JP H0382997 A JPH0382997 A JP H0382997A
- Authority
- JP
- Japan
- Prior art keywords
- uranium
- plutonium
- extraction
- neptunium
- separated
- 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.)
- Granted
Links
- 238000000034 method Methods 0.000 title claims abstract description 24
- 239000002915 spent fuel radioactive waste Substances 0.000 title claims abstract description 6
- 238000012958 reprocessing Methods 0.000 title description 6
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 title 1
- 229910052770 Uranium Inorganic materials 0.000 claims abstract description 52
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 claims abstract description 52
- 229910052778 Plutonium Inorganic materials 0.000 claims abstract description 31
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 claims abstract description 31
- 229910052781 Neptunium Inorganic materials 0.000 claims abstract description 19
- LFNLGNPSGWYGGD-UHFFFAOYSA-N neptunium atom Chemical compound [Np] LFNLGNPSGWYGGD-UHFFFAOYSA-N 0.000 claims abstract description 19
- 238000000638 solvent extraction Methods 0.000 claims abstract description 6
- 229910052768 actinide Inorganic materials 0.000 claims abstract description 3
- 150000001255 actinides Chemical class 0.000 claims abstract description 3
- 230000004992 fission Effects 0.000 claims description 4
- NBIIXXVUZAFLBC-UHFFFAOYSA-N Phosphoric acid Chemical compound OP(O)(O)=O NBIIXXVUZAFLBC-UHFFFAOYSA-N 0.000 claims 2
- 229910000147 aluminium phosphate Inorganic materials 0.000 claims 1
- 239000012074 organic phase Substances 0.000 claims 1
- 238000000605 extraction Methods 0.000 abstract description 41
- 239000003960 organic solvent Substances 0.000 abstract description 15
- 238000000926 separation method Methods 0.000 abstract description 8
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 abstract description 6
- 238000011084 recovery Methods 0.000 abstract description 4
- 230000000694 effects Effects 0.000 abstract description 3
- 239000003795 chemical substances by application Substances 0.000 abstract 1
- 229910017604 nitric acid Inorganic materials 0.000 description 20
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 14
- WYKYKTKDBLFHCY-UHFFFAOYSA-N chloridazon Chemical compound O=C1C(Cl)=C(N)C=NN1C1=CC=CC=C1 WYKYKTKDBLFHCY-UHFFFAOYSA-N 0.000 description 3
- 238000006479 redox reaction Methods 0.000 description 3
- 238000005516 engineering process Methods 0.000 description 2
- 239000000463 material Substances 0.000 description 2
- 239000003758 nuclear fuel Substances 0.000 description 2
- 229910052684 Cerium Inorganic materials 0.000 description 1
- 101000653510 Homo sapiens TATA box-binding protein-like 2 Proteins 0.000 description 1
- 229910019142 PO4 Inorganic materials 0.000 description 1
- 102100030631 TATA box-binding protein-like 2 Human genes 0.000 description 1
- NBSLIVIAZBZBFZ-UHFFFAOYSA-N [Np+4] Chemical compound [Np+4] NBSLIVIAZBZBFZ-UHFFFAOYSA-N 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 239000003153 chemical reaction reagent Substances 0.000 description 1
- 238000005202 decontamination Methods 0.000 description 1
- 230000003588 decontaminative effect Effects 0.000 description 1
- 230000003247 decreasing effect Effects 0.000 description 1
- DPROBIWCCSSTNV-UHFFFAOYSA-N dodecane;tributyl phosphate Chemical compound CCCCCCCCCCCC.CCCCOP(=O)(OCCCC)OCCCC DPROBIWCCSSTNV-UHFFFAOYSA-N 0.000 description 1
- 125000002887 hydroxy group Chemical group [H]O* 0.000 description 1
- 231100001252 long-term toxicity Toxicity 0.000 description 1
- 229910052758 niobium Inorganic materials 0.000 description 1
- UUEIPVJWXIZYGO-UHFFFAOYSA-N nitric acid;tributyl phosphate Chemical compound O[N+]([O-])=O.CCCCOP(=O)(OCCCC)OCCCC UUEIPVJWXIZYGO-UHFFFAOYSA-N 0.000 description 1
- 230000003647 oxidation Effects 0.000 description 1
- 238000007254 oxidation reaction Methods 0.000 description 1
- NBIIXXVUZAFLBC-UHFFFAOYSA-K phosphate Chemical compound [O-]P([O-])([O-])=O NBIIXXVUZAFLBC-UHFFFAOYSA-K 0.000 description 1
- 239000010452 phosphate Substances 0.000 description 1
- 238000004886 process control Methods 0.000 description 1
- 238000000746 purification Methods 0.000 description 1
- 229910052707 ruthenium Inorganic materials 0.000 description 1
- AAORDHMTTHGXCV-UHFFFAOYSA-N uranium(6+) Chemical compound [U+6] AAORDHMTTHGXCV-UHFFFAOYSA-N 0.000 description 1
- 239000002699 waste material Substances 0.000 description 1
- 229910052726 zirconium Inorganic materials 0.000 description 1
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
Landscapes
- 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)
Abstract
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 Field of Application) The present invention provides a method for removing useful nuclear fuel materials such as uranium and plutonium, as well as neptunium, which is important for safety, contained in spent nuclear fuel generated from power reactors, etc. This relates to the sophistication of so-called reprocessing technology, which involves purification, separation and recovery using the Rex method.
（従来の技術）
従来のビ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すいている。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 Burex method reprocessing is basically based on the solvent extraction method at room temperature. However, in extractors that perform redox reactions or reverse extraction of uranium, the temperature is raised 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 （
ＶＩ） 、 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＠） 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 (
VI) and Pu (IV), both of which are well extracted by the above extractant at room temperature. Therefore, when uranium and plutonium are separated from each other, Pu(IV) is reduced to Pu(III), which is difficult to extract, and pu@) is back extracted.
■ ネプツニウムは、液性により（ＩＶ）、 （Ｖ）、
（Ｖｌ）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 classified into (IV), (V),
(Vl) Because the valence and valence of Vl are easily fluctuated and valence adjustment is difficult, no separation and recovery technology has been determined at present, and various methods are being considered.
（発明が解決しようとする課題）
ビ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■価に還元する必要があり
、そのためにＦｅ （ＩＩ） 、Ｕ（ＩＶ） あるい
は硝酸ヒドロキシルア□ン（ＲＡＭ）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。(Problem to be Solved by the Invention) In the Burex method reprocessing, in order to separate uranium and plutonium from each other, it is necessary to reduce the latter to a valence. A chemical reagent such as hydroxyl ane (RAM) is used or an electrochemical reduction method is used.
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皿）をＰｕｆｆＶ） Ｋ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、各元素を高収率、高純度で
分離する事は困難である。These reduction methods not only require complex process controls and equipment, but also require oxidation equipment to reconstitute the The previously known method for recovering neptunium in the reprocessing process attempts to separate uranium, plutonium, and neptunium from each other using a redox reaction in the coexistence of plutonium and neptunium, which are the most reactive. As a result, ■ The redox reaction involving these three elements becomes extremely complex, and it is difficult to separate each element with high yield and high 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、１）抽出系の温度を０〜５°Ｃに低減し、１１）
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, 1) the temperature of the extraction system is reduced to 0 to 5°C; 11)
By increasing the uranium concentration in the organic solvent, the extraction and distribution ratio of plutonium is lowered and the distribution ratio of uranium is increased.By using this effect, both elements can be easily extracted without adjusting or controlling the valence. can be separated. Therefore, troubles caused by erroneous operations can be reduced.
また、本発明では、１ｊ１）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。Furthermore, in the present invention, 1j1) first, plutonium is separated and recovered through a low-temperature, high-load 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、簡便がり、安全に分離・回収出来るもの
である。In other words, the method of the present invention allows actinide elements such as uranium, plutonium, and neptunium in spent nuclear fuel to be easily and safely separated and recovered by reducing the temperature of the extraction process 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・高負荷、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。(Example) 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. The load and extraction method are uranium, plutonium,
We have discovered that this method is extremely effective in separating neptunium from each other.
実施例１
ウラン（Ｖｌ）、プルトニウム曲）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 study the dependence of the distribution ratio of uranium (Vl and plutonium) on temperature and uranium concentration, an extraction operation was performed under the following conditions.
３０ ｖｏ１％ リン酸トリブチル−ドデカンと２−３
Ｍ硝酸抽出系におけるＵ（ＶＩ）とＰｕ（ｒＶ）の分配
比を、口、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、ＰｕＱＶ）の分配
比は減少し、Ｕ（Ｖｌ）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。30 vol1% tributyl phosphate-dodecane and 2-3
The distribution ratio of U (VI) and Pu (rV) in the M nitric acid extraction system was measured under various conditions at 5, 10, 15, 20, and 25 degrees Celsius.The lower the temperature, the lower the distribution ratio of PuQV). It was found that the distribution ratio of U(Vl) decreased and the distribution ratio of U(Vl) increased, and that the higher the uranium concentration in the organic solvent, the lower the distribution ratio of both.
そして、この３０ ｖｏｌ ％リン酸トリブチル（’Ｔ
ＢＰ）−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ＧＶ）、
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。And this 30 vol% tributyl phosphate ('T
BP) - In the nitric acid extraction system, the temperature of the system and the uranium concentration in the organic solvent are uranium (VI), plutonium GV),
Figure 1(a) shows the influence of -
(c).
実施例２
主要な核分裂生成物核種の、高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 study the temperature dependence of the distribution ratio of major fission product nuclides under high uranium loading conditions, an extraction operation was performed under the following conditions.
３０ ｖｏｌ ％リン酸トリブチルードデカンと２．３
Ｍ硝酸抽出系におけるＲｕ、 Ｚｒ、 Ｎｂ、 Ｃｅ
の分配比について、それらの温度依存性を調べた結果、
低温・ウラン高負荷条件下の多段抽出工程では、特にＺ
ｒ、Ｎｂ で顕著な除染係数の向上が認められた。30 vol% tributyldodecane phosphate and 2.3
Ru, Zr, Nb, Ce in M nitric acid extraction system
As a result of investigating the temperature dependence of the distribution ratio of
In multi-stage extraction processes under low temperature and high uranium load conditions, Z
A remarkable improvement in the decontamination coefficient was observed with r and Nb.
そして、このろ［］ｖｏ１％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。Figure 2 shows the dependence of the distribution ratio of the main fission product nuclides on temperature, uranium concentration in the organic solvent, and nitric acid concentration in the Koro[]vo1%TBP-nitric acid extraction system. a) and (b).
実施例ろ
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 Mutual separation of uranium and plutonium by a low-temperature, high-load process was carried out under the extraction process conditions shown in FIG. 3(a).
有機溶媒ＴＢＰ ２００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
、１６０．９／１のウラン及び２ｇ／ｌのプルトニウム
を含有する２Ｍ硝酸溶液ＩＯ［］部が第６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。200 parts of organic solvent TBP are introduced into the first extraction stage of the process, 50 parts of 1M nitric acid solution are introduced into the twelfth extraction stage, and a 2M nitric acid solution containing 160.9/1 uranium and 2 g/l plutonium is added. The IO[ ] portion is introduced into the sixth extraction stage and subjected to countercurrent extraction in a TBP-nitric acid extraction system at an extraction temperature of 5°C.
かかる抽出処理において、８０．９／１のウラン及びろ
Ｘｌ０−’、９／１ のプルトニウムを含有する有機溶
媒抽出系が第１２抽出段から得られ、１．３６ｇ／ｌの
プルトニウム及び０．１３ｇ／ｌのウランを含有する硝
酸溶液抽出系が第１抽出段から得られる。In this extraction process, an organic solvent extraction system containing 80.9/1 uranium and 9/1 plutonium was obtained from the 12th extraction stage, 1.36 g/l plutonium and 0.13 g/l A nitric acid solution extraction system containing uranium/l 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／ｌ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 extracted concentration distribution of uranium, plutonium, etc. obtained as a result is as shown in Figure 3 (b). It was found that uranium products containing plutonium of less than /l can be 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 Mutual separation of uranium and neptunium was carried out under the extraction process conditions shown in FIG. 4(a).
有機溶媒ＴＢＰ２［３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、８５！；／／ｌのウランと０．１．！９／１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．？
／１のウラン（Ｖｌ）、８５ｘ１０−５．９／ｌ及び０
．０２Ｍの硝酸を含有する有機溶媒抽出系が第１２抽出
段から得られ、０．２ｉ／１のネプツニウム（ＩＶ）、
０．００７５ｇ／ｌのウラン及び１．５Ｍの硝酸を含有
する硝酸抽出系が第１抽出段から得られる。3 parts of organic solvent TBP2 were introduced into the first extraction stage of the process, 20 parts of 0.5M nitric acid solution was introduced into the twelfth extraction stage, and 85! ;//l of uranium and 0.1. ! 100 parts of TBP organic solvent containing 9/1 neptunium and 0.02M nitric acid are introduced into the fifth extraction stage, and 50 parts of a 2.5M nitric acid solution are introduced into the sixth extraction stage in which the The flow is extracted and processed. In such an extraction process, 70.8. ?
/1 uranium (Vl), 85x10-5.9/l and 0
．． An organic solvent extraction system containing 0.2M nitric acid is obtained from the 12th extraction stage, 0.2i/1 neptunium(IV),
A nitric acid extraction system containing 0.0075 g/l uranium and 1.5 M 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濃度分
布は第４図（ｂ）に示されるとおりであり、９９９多以
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 extracted concentration distribution of uranium, neptunium, etc. obtained as a result is as shown in FIG. 4(b), and it was found that more than 999% of neptunium was separated and recovered, including 5 wt% or less 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、ビュー
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。(Effects of the Invention) This is produced when a dissolved solution of spent nuclear fuel is processed by the Burex 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、高収率・高純度で分離・回収
することができる。■Uranium and plutonium, which are useful nuclear fuel materials contained in the solution, can be separated and recovered with high yield and purity.
■ 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 simply by lowering the process temperature 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高収率で分離・回収出来
る。■ Uranium and neptunium, which should not be included in waste due to long-term toxicity, can be separated and recovered in high yield.
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（ａＬ （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示すグ
ラフである。
第３図（ａ）は３０ ｖｏｌ係ＴＢＰ−硝酸抽出系の、
低温（５℃）・高負荷Ｕ（ＶＩ） −Ｐｕ ＧＶ）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。Figures 1 (a), (b), and (C) show a 30 vol % tributyl phosphate (TBP)-nitric acid extraction system, where the temperature of the system and the concentration of uranium in the organic solvent are the same as that of uranium (VI) and plutonium (IV). FIG. 2 is a graph showing the influence of Figure 2 (aL) (b) is a graph showing the dependence of the distribution ratio of major fission product nuclides on temperature, uranium concentration in organic solvent, and nitric acid concentration in a 30 vol% TBP-nitric acid extraction system. Figure (a) shows a 30 vol TBP-nitric acid extraction system.
This is a low temperature (5° C.)/high load U(VI)-Pu GV) separation process flow sheet, and FIG. 3(b) is a graph showing the concentration distribution of uranium, plutonium, etc. FIG. 4(a) is a flow sheet for the separation process of uranium and neptunium in a 30 vol% TBP-nitric acid extraction system, and FIG. 4(b) is a graph showing the concentration distribution of uranium, neptunium, etc.
Claims (1)
有機相（抽出剤）飽和度（負荷度）に依存する事を利用
し、低温工程を含む一連の向流多段溶媒抽出工程（リン
酸トリブチル抽出剤）によって、使用済核燃料の溶解液
中に含まれるウラン、プルトニウム及びネプツニウムを
、核分裂生成物から分離すると共に、これらのアクチノ
イド元素を相互に分離回収する方法。Taking advantage of the fact that the distribution ratio of various elements in the solvent extraction system depends on the temperature and the degree of saturation (loading) of the organic phase (extractant), a series of countercurrent multistage solvent extraction steps including a low temperature step (phosphoric acid A method of separating uranium, plutonium, and neptunium contained in a solution of spent nuclear fuel from fission products using a tributyl extractant, and also separating and recovering these actinide elements from each other.
Priority Applications (2)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP21888089A JP2858805B2 (en) | 1989-08-25 | 1989-08-25 | Reprocessing of spent nuclear fuel by low temperature and high load Purex method |
| FR9008286A FR2651364B1 (en) | 1989-08-25 | 1990-06-29 | PROCESS FOR THE RE-TREATMENT OF USED NUCLEAR FUELS. |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP21888089A JP2858805B2 (en) | 1989-08-25 | 1989-08-25 | Reprocessing of spent nuclear fuel by low temperature and high load Purex method |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPH0382997A true JPH0382997A (en) | 1991-04-08 |
| JP2858805B2 JP2858805B2 (en) | 1999-02-17 |
Family
ID=16726752
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP21888089A Expired - Fee Related JP2858805B2 (en) | 1989-08-25 | 1989-08-25 | Reprocessing of spent nuclear fuel by low temperature and high load Purex method |
Country Status (2)
| Country | Link |
|---|---|
| JP (1) | JP2858805B2 (en) |
| FR (1) | FR2651364B1 (en) |
Family Cites Families (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| 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 |
-
1989
- 1989-08-25 JP JP21888089A patent/JP2858805B2/en not_active Expired - Fee Related
-
1990
- 1990-06-29 FR FR9008286A patent/FR2651364B1/en not_active Expired - Fee Related
Also Published As
| Publication number | Publication date |
|---|---|
| JP2858805B2 (en) | 1999-02-17 |
| FR2651364B1 (en) | 1995-01-06 |
| FR2651364A1 (en) | 1991-03-01 |
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