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Influences of the ZrC coating process and heat treatment on ZrC-coated kernels used as fuel in Pu-burner high temperature gas-cooled reactor in Japan

相原 純; 植田 祥平; 本田 真樹*; 水田 直紀; 後藤 実; 橘 幸男; 岡本 孝司*

Journal of Nuclear Science and Technology, 58(1), p.107 - 116, 2021/01

 被引用回数:0 パーセンタイル:100(Nuclear Science & Technology)



炉心溶融物の粘性及び表面張力同時測定技術の開発(委託研究); 令和元年度英知を結集した原子力科学技術・人材育成推進事業

廃炉環境国際共同研究センター; 大阪大学*

JAEA-Review 2020-038, 41 Pages, 2020/12




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.


Study on plutonium burner high temperature gas-cooled reactor in Japan; Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

植田 祥平; 水田 直紀; 深谷 裕司; 後藤 実; 橘 幸男; 本田 真樹*; 齋木 洋平*; 高橋 昌史*; 大平 幸一*; 中野 正明*; et al.

Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02

 被引用回数:1 パーセンタイル:24.17(Nuclear Science & Technology)



炉心溶融物の粘性及び表面張力同時測定技術の開発(委託研究); 平成30年度英知を結集した原子力科学技術・人材育成推進事業

廃炉国際共同研究センター; 大阪大学*

JAEA-Review 2019-025, 36 Pages, 2020/01





相原 純; 後藤 実; 植田 祥平; 橘 幸男

JAEA-Data/Code 2019-018, 22 Pages, 2020/01




Irradiation growth behavior of improved Zr-based alloys for fuel cladding

天谷 政樹; 垣内 一雄; 三原 武

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

New fuel cladding alloys of which composition was changed from conventional ones have been developed by nuclear fuel vendors and utilities. Since the irradiation growth of fuel cladding is one of the most important parameters which determine the dimensional stability of fuel rod and/or fuel assembly during normal operation, the irradiation growth behavior of the improved Zr-based alloys for light-water reactor fuel cladding was investigated. The coupon specimens were prepared from fuel cladding tubes with various kinds of improved Zr-based alloys. The specimens were loaded into test rigs and had been irradiated in the Halden reactor in Norway under several coolant temperature conditions up to a fast-neutron fluence of $$sim$$7.8$$times$$10$$^{21}$$ (n/cm$$^{2}$$, E $$>$$ 1 MeV). Irradiation conditions such as specimen temperatures had been continuously monitored during the irradiation. During and after the irradiation, the amount of irradiation growth of each specimen was evaluated as a part of the interim and final inspections. The effect of the difference in alloy composition on the amount of irradiation growth seemed insignificant if the other conditions e.g. the final heat treatment condition at fabrication and the irradiation temperature were the same.


Structure change of monoclinic ZrO$$_{2}$$ baddeleyite involving softenings of bulk modulus and atom vibrations

福井 宏之*; 藤本 真人*; 赤浜 裕一*; 佐野 亜沙美; 服部 高典

Acta Crystallographica Section B; Structural Science, Crystal Engineering and Materials (Internet), 75(4), p.742 - 749, 2019/08

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



Microstructures of ZrC coated kernels for fuel of Pu-burner high temperature gas-cooled reactor in Japan

相原 純; 植田 祥平; 本田 真樹*; 水田 直紀; 後藤 実; 橘 幸男; 岡本 孝司*

Journal of Nuclear Materials, 522, p.32 - 40, 2019/08



High-temperature interaction between zirconium and UO$$_2$$

白数 訓子; 鈴木 晶大*; 永江 勇二; 倉田 正輝

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05



Dissolution and chemical analysis of Zr-based lanthanide nitrides

林 博和; 千葉 力也*

Progress in Nuclear Science and Technology (Internet), 5, p.196 - 199, 2018/11



Rapid separation of zirconium using microvolume anion-exchange cartridge for $$^{93}$$Zr determination with isotope dilution ICP-MS

浅井 志保; 半澤 有希子; 今田 未来; 鈴木 大輔; 間柄 正明; 木村 貴海; 石原 量*; 斎藤 恭一*; 山田 伸介*; 廣田 英幸*

Talanta, 185, p.98 - 105, 2018/08

 被引用回数:4 パーセンタイル:70.28(Chemistry, Analytical)



A Thermodynamic model for ZrO$$_{2}$$(am) solubility at 25$$^{circ}$$C in the Ca$$^{2+}$$-Na$$^{+}$$-H$$^{+}$$-Cl$$^{-}$$-OH$$^{-}$$-H$$_{2}$$O system; A Critical review

Rai, D.*; 北村 暁; Altmaier, M.*; Rosso, K. M.*; 佐々木 隆之*; 小林 大志*

Journal of Solution Chemistry, 47(5), p.855 - 891, 2018/05

 被引用回数:5 パーセンタイル:85.34(Chemistry, Physical)

ジルコニウムについて、単核および複核の加水分解種の生成定数および非晶質二酸化ジルコニウム(ZrO$$_{2}$$(am))の溶解度積を導出した実験データをレビューした。このレビューを通して、加水分解種(Zr(OH)$$_{2}$$$$^{2+}$$, Zr(OH)$$_{4}$$(aq), Zr(OH)$$_{5}$$$$^{-}$$, Zr(OH)$$_{6}$$$$^{2-}$$およびCa$$_{3}$$Zr(OH)$$_{6}$$$$^{4+}$$)の生成定数やZrO$$_{2}$$(am)の溶解度積を新規に決定もしくは改訂した。


The Influence of the air fraction in steam on the growth of the columnar oxide and the adjacent $$alpha$$-Zr(O) layer on Zry-4 fuel cladding at 1273 and 1473 K

Negyesi, M.; 天谷 政樹

Annals of Nuclear Energy, 114, p.52 - 65, 2018/04

 被引用回数:3 パーセンタイル:40.19(Nuclear Science & Technology)

The growth kinetics of the columnar oxide and $$alpha$$-Zr(O) layers of Zry-4 under mixed steam-air conditions at temperatures of 1273 and 1473 K were investigated in this study be means of post-test metallographic measurements. The hydrogen uptake was also determined by the inert gas fusion technique. The kinetics of the columnar oxide layer obeyed a parabolic law for all air fractions at both temperatures. The kinetics of $$alpha$$-Zr(O) layer appeared to deviate slightly from the parabolic law. The parabolic oxidation rate constant of the columnar oxide increased with increasing air fraction, whereas the parabolic oxidation rate constant of $$alpha$$-Zr(O) layer seemed to be independent of the air fraction. Mixed steam-air conditions appeared to enhance hydrogen absorption substantially, especially after the columnar oxide lost its protectiveness.


余裕深度処分環境におけるふげん圧力管(Zr-2.5wt%Nb合金)の腐食速度の評価,2; 5ヶ年経過データによる長期腐食の考察

菅谷 敏克; 中谷 隆良; 坂井 章浩

JAEA-Technology 2017-032, 21 Pages, 2018/01



Oxidation kinetics of Zry-4 fuel cladding in mixed steam-air atmospheres at temperatures of 1273 - 1473 K

Negyesi, M.; 天谷 政樹

Journal of Nuclear Science and Technology, 54(10), p.1143 - 1155, 2017/10

 被引用回数:5 パーセンタイル:31.96(Nuclear Science & Technology)

This paper deals with the oxidation behavior of Zry-4 nuclear fuel cladding tubes in mixed steam_air atmospheres at temperatures of 1273 and 1473 K. The main goal is to study the oxidation kinetics of Zry-4 fuel cladding in dependence on the air fraction in steam in the range from 0 up to 100%. The purpose of this study is to provide experimental data suitable for an oxidation correlation applicable for thermomechanical analysis codes of nuclear power reactor under severe accidents. The influence of the air addition in steam on parameters of Zry-4 kinetic equation has been quantified using the results of weight gain measurements. At 1273 K, both pre-transient and post-transient regimes were treated. The results of weight gain measurements showed a strong dependence of the Zry-4 oxidation kinetics on the air fraction in steam, especially at 1473 and at 1273 K in the post-transient regime.


Microscopic analyses of complexes formed in adsorbent for Mo and Zr separation chromatography

安倍 諒治*; 名越 航平*; 新井 剛*; 渡部 創; 佐野 雄一; 松浦 治明*; 高木 秀彰*; 清水 伸隆*; 江夏 昌志*; 佐藤 隆博*

Nuclear Instruments and Methods in Physics Research B, 404, p.173 - 178, 2017/08

 被引用回数:1 パーセンタイル:80.48(Instruments & Instrumentation)

Molybdenum and zirconium obstruct the efficient recovery of minor actinides (MA(III): Am(III) and Cm(III)) by extraction chromatography; hence, the removal of these elements prior to MA(III) recovery is desirable. The use of an adsorbent impregnated with bis(2-ethylhexyl) hydrogen phosphate (HDEHP) for Mo and Zr decontamination was evaluated in this report. The adsorption/elution and column separation experiments showed that Mo and Zr in the simulated HLLW were selectively adsorbed on the particles, and that Mo was eluted by H$$_{2}$$O$$_{2}$$. EXAFS analysis and SAXS patterns of the adsorbent containing Zr revealed that the Zr-HDEHP complex had a crystal-like periodic structure similar to the structure of the precipitate produced in the solvent extraction system. Micro-PIXE analysis revealed that distribution of the residual Zr on the adsorbent was uniform.


Extraction and separation of Se, Zr, Pd, and Cs including long-lived radionuclides

佐々木 祐二; 森田 圭介; 鈴木 伸一; 塩飽 秀啓; 伊藤 圭祐; 高橋 優也*; 金子 昌章*

Solvent Extraction Research and Development, Japan, 24(2), p.113 - 122, 2017/06

硝酸溶液からオクタノールまたはドデカン溶媒へのSe, Zr, Pd, Csの溶媒抽出を行った。これら元素は長半減期の核種を含み、高レベル廃液の処理にとってこれら元素の簡便な分離方法の開発が不可欠である。Seはフェニレンジアミン、ZrはHDEHP又はTODGA、PdはMIDOA又はNTAアミドで抽出可能である。CsはDtBuDB18C6を用いて、抽出溶媒を水相の10倍を用いることで90%回収を達成できることを確認した。


Separation of Zr in the rubble waste generated at the Fukushima Daiichi Nuclear Power Station

島田 亜佐子; 亀尾 裕

Journal of Radioanalytical and Nuclear Chemistry, 311(3), p.1613 - 1618, 2017/03


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

福島第一原子力発電所の事故で発生したガレキ中の$$^{93}$$Zrを分析するために、ガレキ中のZrの分離法を開発した。ほぼ100%のZrとNb, Bi, Th, UとMoの一部は3M硝酸溶液からTRUレジンに抽出され、LiやBe, Mg, Al, Ca, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, As, Se, Rb, Sr, Ag, Cd, In, Cs, Baは溶出することでこれらの元素が分離された。ほぼ100%のZrとNb, U、10%のMo、7.1%のHg、77%のBi、20%のThが0.01Mフッ化水素酸により回収された。超寿命核種である$$^{93}$$ZrをICP-MSにより定量するためには、ZrはNbやMoから分離されなければならない。そのため、この回収フラクションを一度乾固した後、0.1Mフッ化水素酸溶液に調製してTEVAレジンに通液し、ZrをNbやMoから分離した。模擬ガレキ試料の溶解液を用いてこの手法の妥当性を検証した。

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