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Microscopic analyses on Zr adsorbed IDA chelating resin by PIXE and EXAFS

荒井 陽一; 渡部 創; 大野 真平; 野村 和則; 中村 文也*; 新井 剛*; 瀬古 典明*; 保科 宏行*; 羽倉 尚人*; 久保田 俊夫*

Nuclear Instruments and Methods in Physics Research B, 477, p.54 - 59, 2020/08

 被引用回数:0 パーセンタイル:100(Instruments & Instrumentation)

Used PUREX process solvent generated from reprocessing process of spent nuclear fuel contains a small amount of U and Pu complexed with tributyl phosphate (TBP) or dibutyl phosphate (DBP). The radioactive nuclides should be removed from the solvent for safety storage or disposal. The iminodiacetic acid (IDA) type chelating resin was proposed as promising procedures for efficient recovery of the trapped cations in the solvent. In order to reveal the distribution and amount of Zr in the particle and local structure of Zr complex formed in the adsorbent, PIXE and EXAFS analyses on the Zr adsorbed chelating resin were carried out. Micro-PIXE analysis proved that it is an effectual method for quantitative analysis of trace adsorbed elements. Moreover, some of the adsorption sites were possibly occupied by the molecules. On the other hand, Zr-K edge EXAFS analysis suggested that extraction mechanism of Zr from the aqueous solution and the solvent was different.


Quantitative analysis of Zr adsorbed on IDA chelating resin using Micro-PIXE

荒井 陽一; 渡部 創; 大野 真平; 野村 和則; 中村 文也*; 新井 剛*; 瀬古 典明*; 保科 宏行*; 久保田 俊夫*

QST-M-23; QST Takasaki Annual Report 2018, P. 59, 2020/03

Radioactive spent solvent waste contains U and Pu is generated from reprocessing process of spent nuclear fuel. The nuclear materials should be removed from the solvent for safety storage or disposal. We are focusing on the nuclear materials recovery from spent solvent using imino diacetic acid (IDA) type chelating resin as a promising method. In order to reveal adsorbed amount of Zr, which is simulated of Pu, Micro-Particle Induced X-ray Emission (PIXE) was carried out. Micro-PIXE analysis succeeded in quantitative analysis on trace amount of adsorbed Zr from simulated spent solvent.


Improvement in flow-sheet of extraction chromatography for trivalent minor actinides recovery

渡部 創; 先崎 達也; 柴田 淳広; 野村 和則; 竹内 正行; 中谷 清治*; 松浦 治明*; 堀内 勇輔*; 新井 剛*

Journal of Radioanalytical and Nuclear Chemistry, 322(3), p.1273 - 1277, 2019/12

 被引用回数:0 パーセンタイル:100(Chemistry, Analytical)

Extraction chromatography flow-sheet employing octyl(phenyl)-$$N,N$$-diisobutylcarbonoylmethylphosphine oxide (CMPO) and $$bis$$(2-ethylhexyl) hydrogen phosphate (HDEHP) extractants for trivalent minor actinide recovery was modified to improve column separation performance. Excellent trivalent minor actinides recovery performance was obtained by column separation experiments on nitric acid solution containing the trivalent minor actinides and representative fission product elements, i.e. recovery yields $$>$$ 93% with sufficient decontamination factors against the fission products. Those are the best performance which we have ever obtained by experiments inside hot cell.


A Review of separation processes proposed for advanced fuel cycles based on technology readiness level assessments

Baron, P.*; Cornet, S. M.*; Collins, E. D.*; DeAngelis, G.*; Del Cul, G.*; Fedorov, Y.*; Glatz, J. P.*; Ignatiev, V.*; 井上 正*; Khaperskaya, A.*; et al.

Progress in Nuclear Energy, 117, p.103091_1 - 103091_24, 2019/11

 被引用回数:3 パーセンタイル:27.1(Nuclear Science & Technology)

本論文では、将来のクローズド燃料サイクルにおける使用済燃料のための分離プロセスに対する国際的リビューの結果が、技術成熟度評価の結果ととともに示されている。本研究は、ORCD/NEAで組織された燃料リサイクル化学に関する専門家グループによって実施されたものである。本研究の特徴的な点は、分離プロセスを使用済燃料中の分離対象元素(ウラン, ウラン-プルトニウム, マイナーアクチノイド, 発熱性元素等)別の分離階層により区分けして評価したことであり、これに使用済燃料の前処理プロセスの評価を加えている。分離プロセスとしては湿式プロセスと乾式プロセスの両者をカバーしている。


Waste management in a hot laboratory of Japan Atomic Energy Agency, 1; Overview and activities in chemical processing facility

野村 和則; 小木 浩通*; 中原 将海; 渡部 創; 柴田 淳広

International Journal of Nuclear and Quantum Engineering (Internet), 13(5), p.209 - 212, 2019/00

Chemical Processing Facility of Japan Atomic Energy Agency is a basic research field for advanced back-end technology developments with using actual high-level radioactive materials. Most of them were treated properly and stored in the liquid waste vessel, but some were not treated and remained at the experimental space as a kind of legacy nuclear waste, which we must treat in safety and dispose if we continue research activities in the facility. Under this circumstance, we launched a collaborative research project called the STRAD project, which stands for Systematic Treatment of Radioactive liquid waste for Decommissioning, in order to develop the treatment processes for wastes of the nuclear research facility. In this project, decomposition methods of certain chemicals, which have been directly solidified without safety pretreatment but may cause a troublesome phenomenon, is developing and a prospect that it will be able to decompose in the facility by simple method. And solidification of aqueous or organic liquid wastes after the decomposition has been studied by adding cement or coagulants. Furthermore, we treated experimental tools of various materials with making an effort to stabilize and to compact them before the package into the waste container. It is expected to decrease the number of transportation of the solid waste and widen the operation space. The project is expected to contribute beneficial waste management outcome that can be shared world widely.


Partitioning of plutonium by acid split method with dissolver solution derived from irradiated fast reactor fuel with high concentration of plutonium

中原 将海; 佐野 雄一; 野村 和則; 竹内 正行

Journal of Chemical Engineering of Japan, 51(3), p.237 - 242, 2018/03

 被引用回数:1 パーセンタイル:87.57(Engineering, Chemical)



Actinides recovery from irradiated fuel for SmART cycle

佐野 雄一; 渡部 創; 中原 将海; 粟飯原 はるか; 竹内 正行

Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 4 Pages, 2017/09



Research of process to treat the radioactive liquid waste containing chloride ion generated by pyroprocessing plant in operating

多田 康平; 北脇 慎一; 渡部 創; 粟飯原 はるか; 柴田 淳広; 野村 和則

Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 3 Pages, 2017/09

塩化物イオン(Cl)を含む放射性廃液は、乾式再処理のプロセス制御における化学分析によって生成される。この廃液を海洋に排出するためには、Clを分離してU, Puを回収する必要がある。本研究では、AgCl沈殿法と抽出クロマトグラフィー法を組み合わせてClを分離し、U, Puを回収した。沈殿試験の結果、UおよびPuが試験後に共沈しないことが分かった。固相抽出試験の結果、95%のPuが液体廃棄物から回収されたことがわかった。Uの濃度が十分でないため、Uについての$$alpha$$放射能を分析することは困難であった。これらの結果は、これらのプロセスが廃液を海に排出する可能性を有することを示した。


日本原子力学会2016秋の大会バックエンド部会企画セッション「ガラス固化体の実力は?; 地層処分におけるガラス固化体性能評価の現状」参加報告

亀井 玄人

原子力バックエンド研究(CD-ROM), 23(2), p.201 - 202, 2016/12



Characterization of the insoluble sludge from the dissolution of irradiated fast breeder reactor fuel

粟飯原 はるか; 荒井 陽一; 柴田 淳広; 野村 和則; 竹内 正行

Procedia Chemistry, 21, p.279 - 284, 2016/12


 被引用回数:3 パーセンタイル:5.95

Insoluble sludge is generated in reprocessing process. Actual sludge data, which had been obtained from the dissolution experiments of irradiated fuel of fast reactor "Joyo" were reevaluated especially from the view point of the characterization of sludge. The yields of sludge were calculated from the weight and there were less than 1%. Element concentrations of sludge were analyzed after decomposing by alkaline fusion. As the results, molybdenum, technetium, ruthenium, rhodium and palladium accounted for mostly of the sludge. From their chemical compositions and structure analyzed by XRD show good agree that main component of sludge is Mo$$_{4}$$Ru$$_{4}$$RhPdTc regardless of the experimental condition. At the condition of reprocessing fast breeder fuel, it is indicated that molybdenum and zirconium in dissolved solution is low, therefore zirconium molybdate hydrate may not produce abundant amount in the process.


Flow-sheet study of MA recovery by extraction chromatography for SmART cycle project

渡部 創; 野村 和則; 北脇 慎一; 柴田 淳広; 小藤 博英; 佐野 雄一; 竹内 正行

Procedia Chemistry, 21, p.101 - 108, 2016/12


 被引用回数:6 パーセンタイル:1.72

Optimization in a flow-sheet of the extraction chromatography process for minor actinides (MA(III); Am and Cm) recovery from high level liquid waste (HLLW) were carried out through batch-wise adsorption/elution experiments on diluted HLLW and column separation experiments on genuine HLLW. Separation experiments using CMPO/SiO$$_{2}$$-P and HDEHP/SiO$$_{2}$$-P adsorbent columns with an improved flow-sheet successfully achieved more than 70 % recovery yields of MA(III) with decontamination factors of Ln(III) $$>$$ 10$$^{3}$$, and a modified flow-sheet for less contamination with fission products was proposed consequently. These results will contribute to MA(III) recovery operations for SmART Cycle project in Japan Atomic Energy Agency which is planned to demonstrates FR fuel cycle with more than 1g of Am.


Simulation study of sludge precipitation in spent fuel reprocessing

竹内 正行; 粟飯原 はるか; 中原 将海; 田中 耕太郎*

Procedia Chemistry, 21, p.182 - 189, 2016/12

 被引用回数:1 パーセンタイル:25.7



Influence of contaminants from spent fuel pools at the Fukushima Daiichi Nuclear Power Station on the reprocessing process

粟飯原 はるか; 北脇 慎一; 野村 和則; 田口 克也

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1076 - 1083, 2015/09

The water in the spent fuel pools at TEPCO Fukushima Daiichi Nuclear Power Station contains sea water and rubbles. When the spent fuels stored in the pools will be reprocessed, it has possibility that these contaminants enter the reprocessing process with the spent fuels. Therefore it is meaningful to estimate the influence of contaminants on reprocessing process in advance. The purpose of this study is to evaluate the behavior and influence of contaminants on the extraction process of spent fuel reprocessing by using simulated contaminants. Contaminants were dissolved into the heated nitric acid and solvent extraction using TBP was performed to obtain distribution ratios. The estimated amount of contaminants accompanied with the spent fuel is low values and solvent extraction tests showed that the distribution ratios of every major element were very low in any case. Also to evaluate the influence of sulfate and chloride ions on uranium and plutonium extraction, Ce(IV) was used for simulated Pu to predict extraction behavior. And then U and Pu test was conducted in order to confirm the simulated test result. Obtained distribution ratio suggests that contaminants will not affect the extraction process.


Chemical composition of insoluble residue generated at the Rokkasho Reprocessing Plant

山岸 功; 小田倉 誠美; 市毛 良明; 黒羽 光彦; 高野 公秀; 赤堀 光雄; 吉岡 正弘*

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1113 - 1119, 2015/09

六ヶ所再処理工場で発生した不溶解残渣の化学組成を分析した。XRD分析では、Mo-Tc-Ru-Rh-Pdからなる白金族合金、モリブデン酸ジルコニウム等の複合Mo酸化物、ジルコニアの存在を確認した。定量した12元素(Ca, Cr, Fe, Ni, Zr, Mo, Tc, Ru, Rh, Pd, Te, U)重量の90%以上は、白金族合金が占めた。シュウ酸溶液で複合Mo酸化物を選択的に洗浄溶解する手法を開発し、白金族合金と複合Mo酸化物の形態で存在するMoの割合を明らかにした。


Study of treatment scenarios for fuel debris removed from Fukushima Daiichi NPS

鷲谷 忠博; 矢野 公彦; 鍛治 直也; 山田 誠也*; 紙谷 正仁

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05



Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

中原 将海; 駒 義和; 中島 靖雄

Journal of Nuclear Science and Technology, 50(11), p.1062 - 1070, 2013/11

 被引用回数:3 パーセンタイル:67.8(Nuclear Science & Technology)

核拡散抵抗性,安全性及びコストに優れたPu還元剤を用いない酸分配法の研究を実施した。抽出計算コードを用いてフローシートの設定を行い、その結果をもとに向流多段抽出試験を行った。Pu逆抽出液は、0.15mol/dm$$^{3}$$ HNO$$_{3}$$を21$$^{circ}$$Cにて供給した。フィード溶液に対してU/Pu製品のPu富化度を2.28倍に高めることができた。また、U製品中におけるPu移行率は、0.47%に抑えられた。本研究により、酸分配法の高速炉燃料再処理への適用性を確認することができた。


Investigation of a LiCl-KCl-UCl$$_{3}$$ system using a combination of X-ray diffraction and differential thermal analyses

仲吉 彬; 北脇 慎一; 福嶋 峰夫; 村上 毅*; 倉田 正輝

Journal of Nuclear Materials, 441(1-3), p.468 - 472, 2013/10

 被引用回数:10 パーセンタイル:28.64(Materials Science, Multidisciplinary)



Plutonium partitioning in uranium and plutonium co-recovery system for fast reactor fuel recycling with enhanced nuclear proliferation resistance

中原 将海; 駒 義和; 中島 靖雄

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.539 - 542, 2013/09



Nitric acid concentration dependence of dicesium plutonium(IV) nitrate formation during solution growth of uranyl nitrate hexahydrate

中原 将海; 鍛治 直也; 矢野 公彦; 柴田 淳広; 竹内 正行; 岡野 正紀; 久野 剛彦

Journal of Chemical Engineering of Japan, 46(1), p.56 - 62, 2013/01

 被引用回数:1 パーセンタイル:92.07(Engineering, Chemical)



Washing of uranyl nitrate hexahydrate crystals with nitric acid aqueous solution to improve crystal quality

中原 将海; 中島 靖雄; 小泉 務

Industrial & Engineering Chemistry Research, 51(46), p.15170 - 15175, 2012/11

 被引用回数:2 パーセンタイル:86.64(Engineering, Chemical)


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