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Ikeuchi, Hirotomo; Koyama, Shinichi; Osaka, Masahiko; Takano, Masahide; Nakamura, Satoshi; Onozawa, Atsushi; Sasaki, Shinji; Onishi, Takashi; Maeda, Koji; Kirishima, Akira*; et al.
JAEA-Technology 2022-021, 224 Pages, 2022/10
JAEA-Technology-2022-021.pdf:12.32MB
A set of technology, including acid dissolving, has to be established for the analysis of content of elements/nuclides in the fuel debris samples. In this project, a blind test was performed for the purpose of clarifying the current level of analytical accuracy and establishing the alternative methods in case that the insoluble residue remains. Overall composition of the simulated fuel debris (homogenized powder having a specific composition) were quantitatively determined in the four analytical institutions in Japan by using their own dissolving and analytical techniques. The merit and drawback for each technique were then evaluated, based on which a tentative flow of the analyses of fuel debris was constructed.
to obtain the high U solution; *; Sakurai, Koji*; ; Nomura, Kazunori; *
JNC TN8400 2000-032, 98 Pages, 2000/12
JNC-TN8400-2000-032.pdf:1.94MB
Concerning the preparation of high U solution for the crystallization process and the application of UO powder dissolution to that, the effects of final U concentration, dissolution temperature, nitric acid concentration and powder size on the dissolution of UO powder in the nitric acid where the final U concentration was 
800g/L were investigated. The experimental results showed that the solubility of UO decreased with the increase of final UO concentration and powder size, and with the decrease of dissolution temperature and nitric acid concentration. It was also confirmed that in the condition where the final U concentration was sufficiently lower than the solubility of U, UO dissolution behavior in the high U solution could be estimated with the equation based on the fragmentation model which we had already reported. Based on these experimental results, the dissolution behavior of irradiated MOX fuel in high U solution was estimated and the possibility of supplying high U solution to the crystallization process was discussed. In the preparation of high U solution for the crystallization process, it was estimated that the present dissolution process (dissolution for fuel pieces of about 3cm long) needed a lot of time to obtain a high dissolution yield, but it was shorted drastically by the pulverization of fuel pieces. The burst of off-gas at the early in the dissolution of fuel powder seems to be avoidable with setting the appropriate dissolution condition, and it is important to optimize the dissolution condition with considering the capacity of off-gas treatment process.
Kihara, Takehiro; *; ; *; *; Fujine, Sachio
JAERI-Research 96-070, 23 Pages, 1997/01
JAERI-Research-96-070.pdf:1.0MB
no abstracts in English
; *; Sugikawa, Susumu; Izawa, Naoki
JAERI-Tech 95-038, 44 Pages, 1995/07
no abstracts in English
Adachi, Takeo; ; *; ; *; Takeishi, Hideyo; Gunji, Katsubumi; Kimura, Takaumi; ; Nakahara, Yoshinori; et al.
Journal of Nuclear Materials, 174, p.60 - 71, 1990/00
Times Cited Count：40 Percentile：94.49(Materials Science, Multidisciplinary)no abstracts in English
; Hagiwara, Miyuki
J.Appl.Polym.Sci., 29, p.965 - 976, 1984/00
Times Cited Count：28 Percentile：79.05(Polymer Science)no abstracts in English
;
JAERI-M 7548, 30 Pages, 1978/02
no abstracts in English
Takeuchi, Masayuki; Miyazaki, Yasunori; Kofuji, Hirohide
no journal, ,
no abstracts in English
Abe, Risako*; Hirasawa, Izumi*; Miyazaki, Yasunori; Takeuchi, Masayuki
no journal, ,
no abstracts in English
Suzuki, Seiya; Arai, Yoichi; Okamura, Nobuo; Watanabe, Masayuki; Kawano, Shohei*; Kawarada, Yoshiyuki*
no journal, ,
The water glass type neutron absorber has been developed as a measure to precaution of re-criticality during fuel debris retrieval. Since the neutron absorber covers the surface of the fuel debris, the drying of the water content of the fuel debris was suggested to be hindered. The drying test using the mock test piece was carried out in order to evaluate the effect on the drying behavior when the surface of the fuel debris is covered with the neutron absorber. We investigate the drying characteristic curve of the mock test piece by thermogravimetric analysis, and report the evaluation of drying behavior.
Arai, Yoichi; Suzuki, Seiya; Okamura, Nobuo; Watanabe, Masayuki; Kawano, Shohei*; Kawaharada, Yoshiyuki*
no journal, ,
The water glass type neutron absorber has been developed as a measure to precaution of re-criticality during fuel debris retrieval from nuclear reactor, etc. Since the surface of the fuel debris is covered with the neutron absorber, the coating of the neutron absorber was suggested to inhibit the evaporation of water in the debris during the drying process of fuel debris. The drying test using the mock test piece was carried out in order to evaluate the effect on the drying behavior when the surface of the fuel debris is covered with the neutron absorber. The outline and basic study of the drying test will be reported in this presentation.
Suzuki, Seiya; Arai, Yoichi; Watanabe, Masayuki; Kawano, Shohei*; Kawaharada, Yoshiyuki*
no journal, ,
The water glass type neutron absorber has been developed as a measure to precaution of re-criticality during fuel debris retrieval at the Fukushima Daiichi Nuclear Power Plant. The drying test using the mock test piece was carried out in order to evaluate the effect on the drying behavior. In this report, the scale-up tests was shown.
Takeuchi, Masayuki
no journal, ,
no abstracts in English
松村 達郎; 津幡 靖宏; 佐野 雄一; 小山 真一
山村 朝雄*; 鷹尾 康一郎*; 鈴木 達也*; 可児 祐子*; 高橋 優也*; 小山 直*; 駒嶺 哲*; 藤田 玲子*; 小澤 正基*
JP, 2018-113698 Patent licensing information Patent publication (In Japanese)
【課題】使用済燃料の再処理工程で発生する不溶解性残渣に含まれるジルコニウムやパラジウムを効率よく回収する方法を提供すること。 【解決手段】不溶解性残渣を過酸化水素含有酸性水溶液等で処理し、不溶解性残渣に含まれるジルコニウムとモリブデンを該酸性水溶液に溶解させて分離する。得られたジルコニウムとモリブデンは、抽出剤などを用いて分離され、ジルコニウムに含まれる長寿命放射性核種は偶奇分離によって低減される。また、酸性水溶液に不溶のパラジウムなどを含む白金族合金については、強酸処理、酸化溶解、フッ素化などによりパラジウムを分離し、パラジウムに含まれる長寿命放射性核種も偶奇分離によって低減される。
