Dissolution behavior of (U,Zr)O-based simulated fuel debris in nitric acid

池内 宏知; 石原 美穂; 矢野 公彦; 鍛治 直也; 中島 靖雄; 鷲谷 忠博

Journal of Nuclear Science and Technology, 51(7-8), p.996 - 1005, 2014/07

https://doi.org/10.1080/00223131.2014.924445

To explore the possibility of dissolving fuel debris as a potential pre-treatment for waste treatment, dissolution tests of UZrO and (U,Pu)ZrO were carried out in 6 M HNO at 353 K. While the U and Zr indicated congruent leaching from the simulated debris with U-rich compositions, a preferential leaching of U was observed with Zr-rich compositions. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). From these findings, dissolution with HNO is expected to be only applicable in U-rich compositions ( 0.3) if the dissolution in 6 M HNO at 353 K is assumed. Application of complexing acids such as mixture of HNO and HF should be considered to increase the dissolution rate of the phases with Zr-rich compositions.

Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi Nuclear Power Station

池内 宏知; 近藤 賀計*; 野口 芳宏*; 矢野 公彦; 鍛治 直也; 鷲谷 忠博

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1349 - 1356, 2013/09

For the decommissioning of Fukushima-Daiichi Nuclear Power Station (1F), characterization of fuel-debris in cores of Unit 1-3 is necessary. In this study, typical phases of fuel-debris generated in reactor pressure vessel were suggested by means of thermodynamic calculation using compositions of core materials and core temperatures. At low ogygen potential where metallic zirconium remains, (U,Zr)O, UO, and ZrO were formed as oxides, and oxygen-dispersed Zr, Fe(Zr,U), and FeUZr were formed as metals. With an increase in zirconium oxidation, the mass of those metals, especially FeUZr, were decreased, but the other phases of metals hardly changed qualitatively. Consequently, (U,Zr)O is suggested as a typical phase of oxide, and Fe(Zr,U) is suggested as that of metal. This result can contribute to the characterization of debris in 1F, which will be also revised by considering the effect of iron content in RPV.

照射済高速中性子炉燃料からのセシウムの水及び希薄硝酸溶液に対する溶出特性

中原 将海; 鍛治 直也; 野村 和則

JAEA-Research 2012-009, 15 Pages, 2012/06

JAEA-Research-2012-009.pdf:6.37MB

晶析工程に関してPuとCsの化合物の生成抑制の観点から、原料液中のCsを減少させることが求められている。照射済核燃料中のCsを粗分離するため、純水及び希薄HNOにおける浸漬試験を行った。浸漬67時間後の純水及び0.1mol/dm HNOにおける燃料粉末からのCsの溶出率は、それぞれ33.8及び38.3%であった。燃料溶解前に燃料粉末を純水もしくは希薄HNO溶液に浸漬させることによりCsを粗分離できる可能性を示した。

A Consideration on proliferation resistance of a FBR fuel cycle system

井上 尚子; 鍛治 直也; 須田 一則; 川久保 陽子; 鈴木 美寿; 小山 智造; 久野 祐輔; 千崎 雅生

Proceedings of INMM 51st Annual Meeting (CD-ROM), 10 Pages, 2010/07

FBR cycle concept has been selected as a promising option for the purpose of the efficient utilization of uranium resources. The innovative technologies and the concern of nuclear proliferation call the needs of the development of advanced safeguards and material accountancy technologies and of the proliferation resistance evaluation methodology that can evaluate as a whole system. In this context, the integrated studies are essential between the system designers and proliferation resistance experts including safeguards. Japan Atomic Energy Agency (JAEA) has started this study for Japanese future nuclear energy system. This is not concluded, however, this joint step-by-step activity provided many benefits and variable findings. This paper describes this activity and its status, and an example of their discussions.

Dissolution of powdered spent fuel and U crystallization from actual dissolver solution for "NEXT" process development

野村 和則; 比内 浩; 中原 将海; 鍛治 直也; 紙谷 正仁; 大山 孝一; 佐野 雄一; 鷲谷 忠博; 小巻 順

Proceedings of 3rd International ATALANTE Conference (ATALANTE 2008) (CD-ROM), 5 Pages, 2008/05

U crystallization process from the dissolver solution of the spent nuclear fuel has been developed as one of the key essential technologies for the "NEXT" process development. Since several tens % of U is supposed to be recovered at the crystallization process, it is expected to reduce the total mass of nuclear material to be treated in the solvent extraction process. For the U crystallization, it is desirable to prepare the dissolver solution of relatively high U concentration. Although the conventional dissolution method needs significantly long dissolution time in order to obtain the dissolver solution of high U concentration with high dissolution ratio, it is expected the effective dissolution is achieved by powderizing the spent fuel. The beaker-scale experiments on the effective powdered fuel dissolution and the U crystallization from dissolver solution with the irradiated MOX fuel from the experimental fast reactor "JOYO" were carried out at the Chemical Processing Facility (CPF) in Tokai Research and Development Center. The powdered fuel was effectively dissolved into the nitric acid solution and the results were compared with the calculation results of the simulation model. In the U crystallization trials, U crystal was obtained from the actual dissolver solution without any addition of reagent.

超臨界流体を用いた全アクチニド一括分離システムの開発,14; 未照射MOX燃料を用いた直接抽出試験

鍛治 直也; 紙谷 正仁; 高畠 容子; 大山 孝一; 三浦 幸子; 小山 智造; 青木 和夫*; 山田 誠也*

no journal, , 

超臨界COを希釈剤に用いる全アクチニド一括抽出システムに関し、未照射MOX燃料を用いた一連の試験を実施した。常圧条件での試験結果に基づく考察から、超臨界条件における初期条件を設定し、さらに超臨界条件での試験結果をもとに条件を微調整することで、U/Pu/Amをほぼ同時に抽出できた。このことから、当初考察に基づく条件設定の妥当性が確認できた。また比較試験により、希釈剤をドデカンに変更してもU/Pu/Amを同時に回収できることを確認した。模擬FPを添加した試験を実施したが、PuもU/Amとともに回収でき、使用済燃料を用いた超臨界直接抽出試験においてみられたPuの残留は再現されなかった。洗浄・逆抽出試験を実施し、従来の抽出プロセスの設計手法の適用により、洗浄,逆抽出工程を設計できる見通しを得た。

溶媒洗浄工程のソルトフリー化で発生する廃液の処理技術開発; 掃気が電解に及ぼす影響の確認

大畠 史一; 大山 孝一; 鍛治 直也; 佐野 雄一

no journal, , 

FaCTプロジェクトでは、湿式再処理工程の一つである溶媒洗浄工程について、廃棄物低減化を目的としたソルトフリー化を検討している。ソルトフリー化した溶媒洗浄工程で発生する洗浄廃液は電解槽による処理を予定されているが、掃気の影響により分解速度が低下する可能性がある。そこで、掃気流量をパラメータとした模擬洗浄廃液の定電流電解を行い、掃気が分解速度に及ぼす影響を評価した。試験結果から分解速度は掃気により阻害されることが示唆されたが、掃気流量の増加に伴う分解速度の大幅な低下は認められなかった。このため、電解時間の延長により目標とする99%以上の分解率を達成できる見込みである。また、気相中に移行する分解生成物を評価した結果からは、施設安全上考慮する必要のあるアジ化水素とアンモニアの移行量は十分に低いことを確認した。

シビアアクシデント後における燃料デブリの湿式処理性に関する検討,2; UO固溶化による燃料デブリの硝酸可溶化処理の提案

矢野 公彦; 星野 貴紀; 鍛治 直也; 鷲谷 忠博; 深澤 哲生*; 小原 浩史*

no journal, , 

福島第一原子力発電所事故において13号機に生成していると考えられる燃料デブリの処置方策の一環として、UOとの固溶化処理により硝酸への溶解を可能とする方法を提案し、デブリの主成分のひとつと考えられる硝酸に難溶な(U,Zr)Oについて、飛躍的に溶解性が向上することを実験的に確認した。

Selection of fuel-debris properties required for defueling work at post severe accident

池内 宏知; 北垣 徹; 涌井 遼平; 樋口 英俊; 小泉 健治; 矢野 公彦; 鍛治 直也; 鷲谷 忠博

no journal, , 

For the decommissioning of Fukushima Daiichi NPP (1F), removal of fuel-debris from the reactor is necessary. In this paper, fuel-debris properties needed for the development of defueling tools were estimated by the following steps. First, the defueling process of Three Mile Island Unit 2 (TMI-2) is summarized briefly. Considering the knowledge of TMI-2 and the tentative results of core-melt analysis of 1F, defueling condition and process for 1F are assumed. Then candidate defueling tools are listed based on expected damaged core condition. Consequently, some physical properties on the fuel-debris which affect applicability of the tools were selected tentatively. In this investigation, mechanical properties such as hardness, elastic modulus and fracture toughness are identified as high priority items to be measured using simulated debris. With reflecting these data, appropriate candidate materials for mock-up test of defueling tools will be designed.

Chemical form estimation of in-vessel fuel debris by thermodynamic calculation with melt progression analysis

池内 宏知; 野口 芳宏*; 近藤 賀計*; 矢野 公彦; 鍛治 直也; 鷲谷 忠博

no journal, , 

For the defueling from Fukushima-Daiichi Nuclear Power Plant (1F) Units 1 to 3, estimation of chemical forms of fuel debris in the reactor pressure vessel (RPV) is necessary. Due to lack of actual information of 1F-core damaged status, the chemical forms had been estimated with the help of thermodynamic equilibration. However, after core degradation, material distribution is quite different from that before the accident. This effect should be considered in the chemical form estimation. In this study, chemical forms of in-vessel fuel debris were roughly evaluated based on the result of melt progression analysis (SA analysis). The case of Unit 2 was especially focused on due to its earliest schedule of defueling.