JP2015125139A - Nuclear fuel substance weighing management method - Google Patents

Nuclear fuel substance weighing management method Download PDF

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JP2015125139A
JP2015125139A JP2013272465A JP2013272465A JP2015125139A JP 2015125139 A JP2015125139 A JP 2015125139A JP 2013272465 A JP2013272465 A JP 2013272465A JP 2013272465 A JP2013272465 A JP 2013272465A JP 2015125139 A JP2015125139 A JP 2015125139A
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separation step
uranium
nuclear fuel
oxide
molten salt
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JP6165623B2 (en
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優也 高橋
Yuya Takahashi
優也 高橋
水口　浩司
Koji Mizuguchi
浩司 水口
中村　等
Hitoshi Nakamura
等 中村
孝 大森
Takashi Omori
孝 大森
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Toshiba Corp
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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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Abstract

PROBLEM TO BE SOLVED: To provide a nuclear fuel substance weighing management method for extracting a nuclear fuel substance mixed in a substance exhibiting insolubility in an acid solution, and appropriately evaluating a total quantity of the extracted nuclear fuel substance.SOLUTION: A nuclear fuel substance weighing management method comprises: a first separation step (S11) of mainly separating metal FE, and metal Zr from fuel debris 10 containing at least UO, (U,Zr)O, the metal Fe, and the metal Zr; a first weighing step (S21) of weighing uranium separated together with the metal Fe and the metal Zr; a second separation step (S12) of immersing a first residue 11 present as a solid in an acid solution, mainly dissolving the UOin the acid solution, and separating the UO; a second weighing step (S22) of weighing the separated uranium; a third separation step (S13) of mainly separating the UOfrom the (U,Zr)Ocontained in a second residue present as a solid; and a third weighing step (S23) of weighing the separated uranium.

Description

本発明は、酸溶液に対し難溶性を示す物質に混在する核燃料物質の計量管理方法に関する。   The present invention relates to a method for measuring and controlling nuclear fuel material mixed in a material that is hardly soluble in an acid solution.

恒常的に軽水炉から発生する使用済燃料は、核燃料である酸化ウランが焼結した状態となっている。
突発的な過酷事故により原子力発電施設で発生する燃料デブリは、酸化ウランを封入した被覆管及び炉内構造物などが溶融し再凝固して形成される。 The spent fuel that is constantly generated from the light water reactor is in a state in which uranium oxide as a nuclear fuel is sintered.
Fuel debris generated in a nuclear power plant due to a sudden severe accident is formed by melting and resolidifying a cladding tube filled with uranium oxide, a reactor internal structure, and the like.

このため燃料デブリは、酸化ウラン、炉内構造物に由来する鉄合金、被覆管に由来するジルコニウム合金、ジルコニウムと水とが高温で反応し生成した酸化ジルコニウム、及びこれら酸化ウランと酸化ジルコニウムとの固溶体が混在した状態となっている。   For this reason, fuel debris is composed of uranium oxide, iron alloys derived from reactor internal structures, zirconium alloys derived from cladding tubes, zirconium oxide produced by the reaction of zirconium and water at high temperatures, and uranium oxide and zirconium oxide. The solid solution is mixed.

核燃料である酸化ウランは、臨界防止の観点から、厳密な計量管理が求められている。
このために、燃料デブリを、原子力発電施設から取り出し中長期的に貯蔵するにあたり、この燃料デブリから酸化ウランを分離回収する必要がある（例えば、特許文献１）。 Uranium oxide, which is a nuclear fuel, requires strict metrological control from the viewpoint of preventing criticality.
For this reason, when the fuel debris is taken out from the nuclear power generation facility and stored for a long period of time, it is necessary to separate and recover uranium oxide from the fuel debris (for example, Patent Document 1).

特開２０１１−２４５２８５号公報JP 2011-245285 A

従来から使用済燃料の再処理に広く用いられている湿式法を燃料デブリに適用した場合、酸溶解により燃料デブリの全量を溶解することは不可能で、酸化ジルコニウムと酸化ウランとの固溶体等といった酸化ウランを含む残渣が発生する。
このように、燃料デブリに含まれる酸化ウランは、硝酸などの酸溶液に対し難溶性を示す上述した物質に混在しているため、分離回収が困難で計量管理に課題があった。 When a wet method that has been widely used for reprocessing spent fuel is applied to fuel debris, it is impossible to dissolve the entire amount of fuel debris by acid dissolution, such as a solid solution of zirconium oxide and uranium oxide, etc. A residue containing uranium oxide is generated.
Thus, since uranium oxide contained in fuel debris is mixed in the above-described substances that are hardly soluble in an acid solution such as nitric acid, separation and recovery are difficult and there is a problem in measurement control.

本発明はこのような事情を考慮してなされたもので、酸溶液に対し難溶性を示す物質に混在する核燃料物質を抽出しその総量を適正に評価する核燃料物質の計量管理方法を提供することを目的とする。   The present invention has been made in view of such circumstances, and provides a method for measuring and controlling nuclear fuel material that extracts nuclear fuel material mixed in a material that is sparingly soluble in an acid solution and appropriately evaluates the total amount thereof. With the goal.

本発明に係る核燃料物質の計量管理方法は、少なくとも酸化ウラン、この酸化ウランと酸化ジルコニウムとの固溶体、金属鉄、及び金属ジルコニウムを包含した燃料デブリから主に前記金属鉄及び前記金属ジルコニウムを分離する第１分離工程と、前記第１分離工程で前記金属鉄及び前記金属ジルコニウムとともに分離されたウランを計量する第１計量工程と、前記第１分離工程の終了後に固体として存在する第１残渣を酸溶液に浸漬して主に前記酸化ウランを溶解させて分離する第２分離工程と、前記第２分離工程で分離された前記ウランを計量する第２計量工程と、前記第２分離工程の終了後に固体として存在する第２残渣に含まれる前記固溶体から主に前記酸化ウランを分離する第３分離工程と、前記第３分離工程で分離された前記ウランを計量する第３計量工程と、を含むことを特徴とする。   The method for measuring and controlling nuclear fuel material according to the present invention mainly separates the metallic iron and the metallic zirconium from at least uranium oxide, a solid solution of the uranium oxide and zirconium oxide, metallic iron, and fuel debris containing metallic zirconium. A first separation step, a first weighing step for weighing uranium separated together with the metallic iron and the metallic zirconium in the first separation step, and a first residue present as a solid after completion of the first separation step A second separation step in which the uranium oxide is mainly dissolved and separated by being immersed in a solution; a second measurement step in which the uranium separated in the second separation step is weighed; and after the end of the second separation step A third separation step for mainly separating the uranium oxide from the solid solution contained in the second residue present as a solid; and the uranium separated in the third separation step. Characterized in that it comprises a third metering step for metering the emission, the.

本発明により、酸溶液に対し難溶性を示す物質に混在する核燃料物質を抽出しその総量を適正に評価する核燃料物質の計量管理方法が提供される。   According to the present invention, there is provided a method for measuring and controlling nuclear fuel material, in which nuclear fuel material mixed in a material that is hardly soluble in an acid solution is extracted and the total amount thereof is appropriately evaluated.

本発明に係る核燃料物質の計量管理方法の第１実施形態を示す工程図。The process figure which shows 1st Embodiment of the measurement management method of the nuclear fuel substance which concerns on this invention. 本発明に係る核燃料物質の計量管理方法の第２実施形態を示す工程図。The process figure which shows 2nd Embodiment of the measurement management method of the nuclear fuel substance which concerns on this invention. 本発明に係る核燃料物質の計量管理方法の第３実施形態を示す工程図。The process figure which shows 3rd Embodiment of the measurement management method of the nuclear fuel substance which concerns on this invention. 本発明に係る核燃料物質の計量管理方法の第４実施形態を示す工程図。Process drawing which shows 4th Embodiment of the measurement management method of the nuclear fuel substance which concerns on this invention. 本発明に係る核燃料物質の計量管理方法の第５実施形態を示す工程図。Process drawing which shows 5th Embodiment of the measurement control method of the nuclear fuel substance which concerns on this invention. 本発明に係る核燃料物質の計量管理方法の第６実施形態を示す工程図。Process drawing which shows 6th Embodiment of the measurement control method of the nuclear fuel substance which concerns on this invention.

（第１実施形態）
以下、本発明の実施形態を添付図面に基づいて説明する。
図１に示すように第１実施形態に係る核燃料物質の計量管理方法は、少なくとも酸化ウラン、この酸化ウランと酸化ジルコニウムとの固溶体、金属鉄及び金属ジルコニウムを包含した燃料デブリ１０から主に金属鉄及び金属ジルコニウム（Ｆｅ，Ｚｒ）を分離する第１分離工程（Ｓ１１）と、この第１分離工程（Ｓ１１）で金属鉄及び金属ジルコニウム（Ｆｅ，Ｚｒ）とともに分離されたウラン（Ｕ）を計量する第１計量工程（Ｓ２１）と、第１分離工程（Ｓ１１）の終了後に固体として存在する第１残渣１１を酸溶液に浸漬して主に酸化ウラン（ＵＯ2）を溶解させて分離する第２分離工程（Ｓ１２）と、この第２分離工程（Ｓ１２）で分離されたウラン（Ｕ）を計量する第２計量工程（Ｓ２２）と、この第２分離工程（Ｓ１２）の終了後に固体として存在する第２残渣１２に含まれる固溶体（(U,Zr)O2）から主に酸化ウラン（ＵＯ2）を分離する第３分離工程（Ｓ１３）と、この第３分離工程（Ｓ１３）で分離されたウラン（Ｕ）を計量する第３計量工程（Ｓ２３）と、を含んでいる。 (First embodiment)
Hereinafter, embodiments of the present invention will be described with reference to the accompanying drawings.
As shown in FIG. 1, the nuclear fuel material measurement control method according to the first embodiment mainly includes metallic iron from a fuel debris 10 including at least uranium oxide, a solid solution of the uranium oxide and zirconium oxide, metallic iron and metallic zirconium. And a first separation step (S11) for separating metallic zirconium (Fe, Zr), and uranium (U) separated together with metallic iron and metallic zirconium (Fe, Zr) in this first separation step (S11) are weighed After the first metering step (S21) and the first separation step (S11), the first residue 11 present as a solid is immersed in an acid solution to mainly dissolve and separate uranium oxide (UO 2 ). A separation step (S12), a second weighing step (S22) for weighing uranium (U) separated in the second separation step (S12), and after the end of the second separation step (S12) Solid solution contained in the second residue 12 present as a solid ((U, Zr) O 2 ) mainly from uranium oxide third separation step of separating (UO 2) and (S13), the third separation step (S13) And a third measuring step (S23) for measuring the uranium (U) separated in (1).

燃料デブリ１０には、核燃料である酸化ウラン（ＵＯ2）を主体として、炉内構造物に由来する金属鉄（Ｆｅ）、被覆管に由来する金属ジルコニウム（Ｚｒ）、ジルコニウムと水とが高温で反応し生成した酸化ジルコニウム（ＺｒＯ2）、及びこれら酸化ウランと酸化ジルコニウムとの固溶体（(U,Zr)O2）が混在した状態となっている。
第１分離工程（Ｓ１１）は、酸溶液に対して難溶性を示す金属鉄及び金属ジルコニウム（Ｆｅ，Ｚｒ）を燃料デブリ１０から分離することを目的としている。 The fuel debris 10 is mainly composed of nuclear fuel uranium oxide (UO 2 ), metallic iron (Fe) derived from the reactor internal structure, metallic zirconium (Zr) derived from the cladding tube, zirconium and water at high temperatures. Zirconium oxide (ZrO 2 ) produced by reaction and a solid solution ((U, Zr) O 2 ) of these uranium oxide and zirconium oxide are mixed.
The first separation step (S11) is intended to separate metallic iron and metallic zirconium (Fe, Zr), which are hardly soluble in an acid solution, from the fuel debris 10.

第１実施形態における第１分離工程（Ｓ１１）は、燃料デブリ１０を陽極で支持し溶融塩に浸漬し電解して陰極に金属鉄及び金属ジルコニウム（Ｆｅ，Ｚｒ）を析出させる。
このように、Ｆｅ，Ｚｒを燃料デブリ１０から分離する第１分離工程（Ｓ１１）の他の実施例として、燃料デブリ１０を液体金属に浸漬して金属鉄及び金属ジルコニウム（Ｆｅ，Ｚｒ）を溶解させる方法がある。そのような液体金属としては、Ｆｅ，Ｚｒと相溶する水銀、カドミウム等が挙げられる。 In the first separation step (S11) in the first embodiment, the fuel debris 10 is supported by an anode, immersed in a molten salt and electrolyzed to deposit metallic iron and metallic zirconium (Fe, Zr) on the cathode.
As described above, as another example of the first separation step (S11) for separating Fe and Zr from the fuel debris 10, the fuel debris 10 is immersed in a liquid metal to dissolve metal iron and metal zirconium (Fe, Zr). There is a way to make it. Examples of such a liquid metal include mercury and cadmium that are compatible with Fe and Zr.

第１計量工程（Ｓ２１）は、第１分離工程（Ｓ１１）において、溶融塩の浴中にＦｅ，Ｚｒとともに分離したウランや、陰極にＦｅ，Ｚｒとともに析出したウランを計量する。
具体的には、ＩＣＰ質量分析装置（ＩＣＰ−ＭＳ）等といった多元素の定性定量分析・同位体分析が可能な分析装置により、ウラン濃度を測定し、分離されたウランの総量を計量する。
第１分離工程（Ｓ１１）で燃料デブリ１０から分離した酸化ウランは、溶融塩や電極析出金属に均一に分散しているので、少量を採取するだけで、臨界防止のための計量管理を厳密に行うことができる。 In the first weighing step (S21), uranium separated together with Fe and Zr in the molten salt bath and uranium precipitated together with Fe and Zr in the molten salt bath in the first separation step (S11) are weighed.
Specifically, the uranium concentration is measured by an analyzer capable of qualitative quantitative analysis / isotope analysis of multiple elements such as an ICP mass spectrometer (ICP-MS), and the total amount of separated uranium is measured.
Since the uranium oxide separated from the fuel debris 10 in the first separation step (S11) is uniformly dispersed in the molten salt and the electrode deposit metal, the measurement control for preventing criticality can be strictly performed by collecting a small amount. It can be carried out.

第１分離工程（Ｓ１１）の終了後には、酸化ウラン（ＵＯ2）、酸化ジルコニウム（ＺｒＯ2）、及びこれらの固溶体（(U,Zr)O2）を主成分とする第１残渣１１が固体として存在している。
第２分離工程（Ｓ１２）では、第１残渣１１を酸溶液に浸漬して主に酸化ウラン（ＵＯ2）を溶解させて分離する。この場合の酸溶液は硝酸が好適に採用されるが、酸化ウラン（ＵＯ2）を溶解させる機能を有するものであれば適宜用いることができる。 After the completion of the first separation step (S11), the first residue 11 mainly composed of uranium oxide (UO 2 ), zirconium oxide (ZrO 2 ), and their solid solution ((U, Zr) O 2 ) is solid. Exist as.
In the second separation step (S12), the first residue 11 is immersed in an acid solution, and mainly uranium oxide (UO 2 ) is dissolved and separated. Nitric acid is preferably used as the acid solution in this case, but any acid solution can be used as long as it has a function of dissolving uranium oxide (UO 2 ).

この第１残渣１１は、酸溶液に難溶性を示す金属鉄及び金属ジルコニウムが除去されているので、含まれる酸化ウラン（ＵＯ2）が効率良く酸溶液に溶解する。
しかし、酸化ジルコニウムとの固溶体（(U,Zr)O2）に含まれる酸化ウラン（ＵＯ2）は、この酸溶液に対し、難溶性を示す。 In the first residue 11, metal iron and metal zirconium which are hardly soluble in the acid solution are removed, so that uranium oxide (UO 2 ) contained therein is efficiently dissolved in the acid solution.
However, uranium oxide (UO 2 ) contained in a solid solution with zirconium oxide ((U, Zr) O 2 ) exhibits poor solubility in this acid solution.

第２分離工程（Ｓ１２）は、図示を省略するが、酸溶液に溶解した酸化ウラン（ＵＯ2）を、さらに溶媒抽出法、沈殿法及びイオン交換法のうちいずれかを用いて酸化ウラン（ＵＯ2）を固相として回収する場合もある。
特にシュウ酸もしくはＮ−シクロヘキシル−２−ピロリドンなどを用いた沈殿法によれば、酸化ウラン（ＵＯ2）の回収方法を簡素化させ、使用する装置の小型化を図ることができる。 In the second separation step (S12), although not shown, uranium oxide (UO 2 ) dissolved in an acid solution is further converted into uranium oxide (UO 2 ) using any one of a solvent extraction method, a precipitation method, and an ion exchange method. 2 ) may be recovered as a solid phase.
In particular, according to the precipitation method using oxalic acid or N-cyclohexyl-2-pyrrolidone, the method for recovering uranium oxide (UO 2 ) can be simplified and the apparatus used can be downsized.

第２計量工程（Ｓ２２）は、この第２分離工程（Ｓ１２）で、酸溶液に溶解している酸化ウラン（ＵＯ2）を計量する。
第２分離工程（Ｓ１２）で第１残渣１１から分離した酸化ウランは、酸溶液に均一に分散しているので、少量を採取するだけで、臨界防止のための計量管理を厳密に行うことができる。
なお、溶媒抽出法、沈殿法又はイオン交換法等によって固相分離された酸化ウラン（ＵＯ2）についても同様である。 In the second measurement step (S22), uranium oxide (UO 2 ) dissolved in the acid solution is measured in the second separation step (S12).
Since the uranium oxide separated from the first residue 11 in the second separation step (S12) is uniformly dispersed in the acid solution, it is possible to strictly control the measurement for preventing criticality by collecting a small amount. it can.
The same applies to uranium oxide (UO 2 ) separated by solid phase by a solvent extraction method, a precipitation method, an ion exchange method or the like.

第２分離工程（Ｓ１２）の終了後には、酸化ジルコニウム（ＺｒＯ2）、及び酸化ウランと酸化ジルコニウムとの固溶体（(U,Zr)O2）を主成分とする第２残渣１２が固体として存在している。 After the completion of the second separation step (S12), zirconium oxide (ZrO 2 ) and a second residue 12 mainly composed of a solid solution of uranium oxide and zirconium oxide ((U, Zr) O 2 ) are present as a solid. doing.

第３分離工程（Ｓ１３）は、第２残渣１２を溶融塩に浸漬し固溶体（(U,Zr)O2）から酸化ウラン（ＵＯ2）を溶解して分離する。そのような溶融塩としては、酸化物系の溶融塩が採用され、特にモリブデン酸溶融塩が好適である。
なお溶融塩による固溶体（(U,Zr)O2）の溶解は、溶解条件の設定により酸化ウラン（ＵＯ2）を選択的に溶解させたり、酸化ジルコニウム（ＺｒＯ2）も含めた全量を溶解させたりもできる。 In the third separation step (S13), the second residue 12 is immersed in a molten salt to dissolve and separate uranium oxide (UO 2 ) from the solid solution ((U, Zr) O 2 ). As such a molten salt, an oxide-based molten salt is employed, and a molybdic acid molten salt is particularly preferable.
The dissolution of the solid solution ((U, Zr) O 2 ) with molten salt can be performed by selectively dissolving uranium oxide (UO 2 ) or by dissolving the entire amount including zirconium oxide (ZrO 2 ) by setting the dissolution conditions. You can also.

第３計量工程（Ｓ２３）は、この第３分離工程（Ｓ１３）で溶融塩に溶解している酸化ウラン（ＵＯ2）を計量する。
第３分離工程（Ｓ１３）で、第２残渣１２から分離した酸化ウランは、溶融塩に均一に分散しているので、少量を採取するだけで、臨界防止のための計量管理を厳密に行うことができる。
第３分離工程（Ｓ１３）の終了後には、酸化ジルコニウム（ＺｒＯ2）を主成分とする第３残渣１３が固体として存在している。 In the third measurement step (S23), uranium oxide (UO 2 ) dissolved in the molten salt in the third separation step (S13) is measured.
Since the uranium oxide separated from the second residue 12 in the third separation step (S13) is uniformly dispersed in the molten salt, the metric management for preventing criticality should be strictly performed by collecting a small amount. Can do.
After the completion of the third separation step (S13), the third residue 13 containing zirconium oxide (ZrO 2 ) as a main component exists as a solid.

このように、第１分離工程（Ｓ１１）、第２分離工程（Ｓ１２）及び第３分離工程（Ｓ１３）を経ることにより、燃料デブリ１０に含まれる酸化ウランを分離し、他の媒体に対して均一に分散させた状態で計量管理することができるため、臨界が防止された保存を行うことができる。   As described above, the uranium oxide contained in the fuel debris 10 is separated by passing through the first separation step (S11), the second separation step (S12), and the third separation step (S13). Since the metering can be performed in a uniformly dispersed state, it is possible to perform storage in which criticality is prevented.

（第２実施形態）
次に図２に基づいて本発明における第２実施形態について説明する。
なお、図２において図１と共通の構成又は機能を有する部分は、同一符号で示し、重複する説明を省略する。
第２実施形態の核燃料物質の計量管理方法では、第３分離工程（Ｓ１３）において、酸化モリブデン及び酸化タングステンのうち少なくとも一つが溶融助剤１４として溶融塩に投入される。
モリブデン酸溶融塩の場合は酸化モリブデン（ＭｏＯ3）が、塩化物系溶融塩の場合は塩化モリブデン（ＭｏＣｌ6）が、溶融助剤１４として好適に用いられる。
これにより、固溶体（(U,Zr)O2）に含まれる酸化ウラン（ＵＯ2）の溶融塩への溶解を促進させることができる。 (Second Embodiment)
Next, a second embodiment of the present invention will be described based on FIG.
2 that have the same configuration or function as those in FIG. 1 are denoted by the same reference numerals, and redundant description is omitted.
In the method for measuring and controlling nuclear fuel material according to the second embodiment, in the third separation step (S13), at least one of molybdenum oxide and tungsten oxide is added as molten auxiliary agent 14 to the molten salt.
Molybdenum oxide (MoO 3 ) is preferably used as the melting aid 14 in the case of molybdic acid molten salt and molybdenum chloride (MoCl 6 ) in the case of chloride-based molten salt.
Thereby, dissolution of the uranium oxide (UO 2 ) contained in the solid solution ((U, Zr) O 2 ) into the molten salt can be promoted.

（第３実施形態）
次に図３に基づいて本発明における第３実施形態について説明する。
なお、図３において図１と共通の構成又は機能を有する部分は、同一符号で示し、重複する説明を省略する。
第３実施形態の核燃料物質の計量管理方法では、第３分離工程（Ｓ１３）において、溶融塩は塩化物系の溶融塩であり塩素ガスが吹き込まれている。
これにより、溶融塩に浸漬された固溶体（(U,Zr)O2）に含まれる酸化ウラン（ＵＯ2）の溶解を促進させることができる。 (Third embodiment)
Next, a third embodiment of the present invention will be described based on FIG.
3, parts having the same configuration or function as those in FIG. 1 are denoted by the same reference numerals, and redundant description is omitted.
In the method for measuring and controlling nuclear fuel material according to the third embodiment, in the third separation step (S13), the molten salt is a chloride-based molten salt, and chlorine gas is injected.
Thereby, dissolution of uranium oxide (UO 2 ) contained in the solid solution ((U, Zr) O 2 ) immersed in the molten salt can be promoted.

（第４実施形態）
次に図４に基づいて本発明における第４実施形態について説明する。
なお、図４において図１と共通の構成又は機能を有する部分は、同一符号で示し、重複する説明を省略する。
第４実施形態の核燃料物質の計量管理方法では、第３分離工程（Ｓ１３）において、溶融塩溶解させた後（Ｓ１３Ａ）、さらに溶融塩に電極を挿入し溶解している酸化ウラン（ＵＯ2）を電極に析出させて回収する（Ｓ１３Ｂ）。
これにより、酸化ウラン（ＵＯ2）を固相として回収することができる。 (Fourth embodiment)
Next, a fourth embodiment of the present invention will be described based on FIG.
4, parts having the same configuration or function as those in FIG. 1 are denoted by the same reference numerals, and redundant description is omitted.
In the method for measuring and controlling nuclear fuel material according to the fourth embodiment, in the third separation step (S13), after melting the molten salt (S13A), an electrode is inserted into the molten salt and dissolved (UO 2 ). Is deposited on the electrode and collected (S13B).
Thereby, uranium oxide (UO 2 ) can be recovered as a solid phase.

（第５実施形態）
次に図５に基づいて本発明における第５実施形態について説明する。
なお、図５において図１と共通の構成又は機能を有する部分は、同一符号で示し、重複する説明を省略する。
第５実施形態の核燃料物質の計量管理方法では、第３分離工程（Ｓ１３）において、第２残渣１２を電解還元して酸化ウラン（ＵＯ2）を金属ウラン（Ｕ）に変換し（Ｓ１３Ｃ）、この金属ウラン（Ｕ）を溶融塩電解により電極に析出させる（Ｓ１３Ｄ）。
これにより、金属ウラン（Ｕ）を固相として回収することができる。 (Fifth embodiment)
Next, a fifth embodiment of the present invention will be described with reference to FIG.
5 that have the same configuration or function as those in FIG. 1 are denoted by the same reference numerals, and redundant description is omitted.
In the nuclear fuel material measurement control method of the fifth embodiment, in the third separation step (S13), the second residue 12 is electrolytically reduced to convert uranium oxide (UO 2 ) to metal uranium (U) (S13C), This metal uranium (U) is deposited on the electrode by molten salt electrolysis (S13D).
Thereby, metal uranium (U) can be recovered as a solid phase.

（第６実施形態）
次に図６に基づいて本発明における第６実施形態について説明する。
なお、図６において図１と共通の構成又は機能を有する部分は、同一符号で示し、重複する説明を省略する。
第６実施形態の核燃料物質の計量管理方法では、第３分離工程（Ｓ１３）において、第２残渣１２にフッ素を反応させて生成したフッ化ウラン（ＵＦ6）を揮発させて（Ｓ１３Ｅ）、冷却トラップにより回収する（Ｓ１３Ｆ）。
固溶体（(U,Zr)O2）にフッ素と反応させることにより、揮発性のフッ化ウラン（ＵＦ6）、フッ化ジルコニウム（ＺｒＦ4）を揮発させて冷却トラップで回収することができる。
この際に、温度差を利用して、フッ化ウラン（ＵＦ6）を選択的に回収することもできる。 (Sixth embodiment)
Next, a sixth embodiment of the present invention will be described based on FIG.
In FIG. 6, parts having the same configuration or function as those in FIG. 1 are denoted by the same reference numerals, and redundant description is omitted.
In the nuclear fuel material measurement control method of the sixth embodiment, in the third separation step (S13), uranium fluoride (UF 6 ) generated by reacting fluorine with the second residue 12 is volatilized (S13E), and cooling is performed. It collects with a trap (S13F).
By reacting solid solution ((U, Zr) O 2 ) with fluorine, volatile uranium fluoride (UF 6 ) and zirconium fluoride (ZrF 4 ) can be volatilized and recovered by a cooling trap.
At this time, uranium fluoride (UF 6 ) can be selectively recovered using the temperature difference.

以上述べた少なくともひとつの実施形態の核燃料物質の計量管理方法によれば、酸溶液に難溶性を示す金属鉄、金属ジルコニウム、及び酸化ウランと酸化ジルコニウムとの固溶体を含む燃料デブリから、核燃料物質である酸化ウランを抽出し、その総量を適正に評価することができる。   According to the method for measuring and controlling nuclear fuel material according to at least one embodiment described above, nuclear fuel material is obtained from fuel debris containing metallic iron, metal zirconium, and a solid solution of uranium oxide and zirconium oxide that are hardly soluble in an acid solution. A certain amount of uranium oxide can be extracted and its total amount can be evaluated appropriately.

本発明のいくつかの実施形態を説明したが、これらの実施形態は、例として提示したものであり、発明の範囲を限定することは意図していない。これら実施形態は、その他の様々な形態で実施されることが可能であり、発明の要旨を逸脱しない範囲で、種々の省略、置き換え、変更、組み合わせを行うことができる。これら実施形態やその変形は、発明の範囲や要旨に含まれると同様に、特許請求の範囲に記載された発明とその均等の範囲に含まれるものである。   Although several embodiments of the present invention have been described, these embodiments are presented by way of example and are not intended to limit the scope of the invention. These embodiments can be implemented in various other forms, and various omissions, replacements, changes, and combinations can be made without departing from the scope of the invention. These embodiments and their modifications are included in the scope and gist of the invention, and are also included in the invention described in the claims and the equivalents thereof.

１０…燃料デブリ、１１…第１残渣、１２…第２残渣、１３…第３残渣、１４…溶融助剤、Ｓ１１…第１分離工程、Ｓ１２…第２分離工程、Ｓ１３（Ｓ１３Ａ、Ｓ１３Ｂ、Ｓ１３Ｃ、Ｓ１３Ｄ、Ｓ１３Ｅ、Ｓ１３Ｆ）…第３分離工程、Ｓ２１…第１計量工程、Ｓ２２…第２計量工程、Ｓ２３…第３計量工程。   DESCRIPTION OF SYMBOLS 10 ... Fuel debris, 11 ... 1st residue, 12 ... 2nd residue, 13 ... 3rd residue, 14 ... Melting aid, S11 ... 1st separation process, S12 ... 2nd separation process, S13 (S13A, S13B, S13C , S13D, S13E, S13F) ... third separation step, S21 ... first measurement step, S22 ... second measurement step, S23 ... third measurement step.

Claims (10)

少なくとも酸化ウラン、この酸化ウランと酸化ジルコニウムとの固溶体、金属鉄、及び金属ジルコニウムを包含した燃料デブリから主に前記金属鉄及び前記金属ジルコニウムを分離する第１分離工程と、
前記第１分離工程で前記金属鉄及び前記金属ジルコニウムとともに分離されたウランを計量する第１計量工程と、
前記第１分離工程の終了後に固体として存在する第１残渣を酸溶液に浸漬して主に前記酸化ウランを溶解させて分離する第２分離工程と、
前記第２分離工程で分離された前記ウランを計量する第２計量工程と、
前記第２分離工程の終了後に固体として存在する第２残渣に含まれる前記固溶体から主に前記酸化ウランを分離する第３分離工程と、
前記第３分離工程で分離された前記ウランを計量する第３計量工程と、を含むことを特徴とする核燃料物質の計量管理方法。 A first separation step of separating mainly the metallic iron and the metallic zirconium from at least uranium oxide, a solid solution of the uranium oxide and zirconium oxide, metallic iron, and fuel debris containing the metallic zirconium;
A first metering step of weighing the uranium separated together with the metal iron and the metal zirconium in the first separation step;
A second separation step of immersing a first residue present as a solid after completion of the first separation step in an acid solution to mainly dissolve and separate the uranium oxide;
A second weighing step for weighing the uranium separated in the second separation step;
A third separation step for mainly separating the uranium oxide from the solid solution contained in the second residue present as a solid after the completion of the second separation step;
And a third measuring step for measuring the uranium separated in the third separating step. 前記第１分離工程は、前記燃料デブリを陽極で支持し溶融塩に浸漬し電解して陰極に前記金属鉄及び前記金属ジルコニウムを析出させることを特徴とする請求項１に記載の核燃料物質の計量管理方法。   2. The measurement of nuclear fuel material according to claim 1, wherein in the first separation step, the fuel debris is supported by an anode, immersed in a molten salt, and electrolyzed to deposit the metallic iron and the metallic zirconium on the cathode. Management method. 前記第１分離工程は、前記燃料デブリを液体金属に浸漬して前記金属鉄及び前記金属ジルコニウムを溶解させることを特徴とする請求項１に記載の核燃料物質の計量管理方法。   The method of claim 1, wherein the first separation step comprises immersing the fuel debris in a liquid metal to dissolve the metallic iron and the metallic zirconium. 前記第２分離工程において、さらに溶媒抽出法、沈殿法及びイオン交換法のうちいずれかを用いて酸化ウランを回収することを特徴とする請求項１から請求項３のいずれか１項に記載の核燃料物質の計量管理方法。   4. The uranium oxide is further recovered in any one of a solvent extraction method, a precipitation method, and an ion exchange method in the second separation step. 5. Measurement method of nuclear fuel material. 前記第３分離工程は、前記第２残渣を溶融塩に浸漬し前記固溶体から前記酸化ウランを溶解させて分離することを特徴とする請求項１から請求項４のいずれか１項に記載の核燃料物質の計量管理方法。   5. The nuclear fuel according to claim 1, wherein, in the third separation step, the second residue is immersed in a molten salt to separate the uranium oxide from the solid solution. Method for measuring and controlling substances. 前記第３分離工程において、酸化モリブデン及び酸化タングステンのうち少なくとも一つを溶融助剤として前記溶融塩に投入することを特徴とする請求項５に記載の核燃料物質の計量管理方法。   6. The method for measuring and controlling a nuclear fuel material according to claim 5, wherein in the third separation step, at least one of molybdenum oxide and tungsten oxide is added to the molten salt as a melting aid. 前記第３分離工程において、前記溶融塩は塩化物系の溶融塩であり塩素ガスが吹き込まれることを特徴とする請求項５又は請求項６に記載の核燃料物質の計量管理方法。   The method for measuring and controlling a nuclear fuel material according to claim 5 or 6, wherein, in the third separation step, the molten salt is a chloride-based molten salt, and chlorine gas is injected. 前記第３分離工程において、さらに前記溶融塩に電極を挿入し溶解している前記酸化ウランを電極に析出させて回収することを特徴とする請求項５から請求項７のいずれか１項に記載の核燃料物質の計量管理方法。   8. The method according to claim 5, wherein, in the third separation step, the uranium oxide dissolved by inserting an electrode into the molten salt is further deposited and recovered on the electrode. Management method for nuclear fuel materials. 前記第３分離工程は、前記第２残渣を電解還元して前記酸化ウランを金属ウランに変換しこの金属ウランを溶融塩電解により電極に析出させることを特徴とする請求項１から請求項４のいずれか１項に記載の核燃料物質の計量管理方法。   5. The method according to claim 1, wherein in the third separation step, the second residue is electrolytically reduced to convert the uranium oxide into metal uranium, and the metal uranium is deposited on an electrode by molten salt electrolysis. The nuclear fuel material measurement management method according to any one of the above. 前記第３分離工程は、前記第２残渣にフッ素を反応させて生成したフッ化ウランを揮発させて冷却トラップにより回収することを特徴とする請求項１から請求項４のいずれか１項に記載の核燃料物質の計量管理方法。   5. The third separation step according to claim 1, wherein in the third separation step, uranium fluoride generated by reacting fluorine with the second residue is volatilized and recovered by a cooling trap. 6. Management method for nuclear fuel materials.
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