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Journal Articles

STRAD project for systematic treatments of radioactive liquid wastes generated in nuclear facilities

Watanabe, So; Ogi, Hiromichi*; Arai, Yoichi; Aihara, Haruka; Takahatake, Yoko; Shibata, Atsuhiro; Nomura, Kazunori; Kamiya, Yuichi*; Asanuma, Noriko*; Matsuura, Haruaki*; et al.

Progress in Nuclear Energy, 117, p.103090_1 - 103090_8, 2019/11

AA2019-0193.pdf:1.29MB

 Times Cited Count:6 Percentile:75.46(Nuclear Science & Technology)

Journal Articles

Applicability of polyvinylpolypyrrolidone adsorbent to treatment process of wastes containing uranium

Ohashi, Yusuke; Harada, Masayuki*; Asanuma, Noriko*; Ando, Shion; Tanaka, Yoshio; Ikeda, Yasuhisa*

Journal of Radioanalytical and Nuclear Chemistry, 311(1), p.491 - 502, 2017/01

 Times Cited Count:1 Percentile:15.39(Chemistry, Analytical)

In order to assess the feasibility of method for recovering U from wastes containing uranium (scrap uranium) using polyvinylpolypyrrolidone (PVPP) adsorbent, we have examined the adsorption and desorption behavior of metal species in HCl aqueous solutions dissolving scrap uranium. It was found that the U(VI) species are selectively adsorbed onto PVPP regardless of the presence of a large amount of Na(I) and Al(III), that the adsorbed U(VI) species are desorbed from PVPP column selectively by water. Pure uranium was efficiently recovered from the eluates. From these results, the PVPP resin is expected to be used as the adsorbent in the treatment process of scrap uranium.

Journal Articles

Studies on electrochemical behavior of uranium species in choline chloride-urea eutectic for developing electrolytically treating method of uranium-bearing wastes

Ohashi, Yusuke; Asanuma, Noriko*; Harada, Masayuki*; Tanaka, Yoshio; Ikeda, Yasuhisa*

Journal of Radioanalytical and Nuclear Chemistry, 309(2), p.627 - 636, 2016/08

 Times Cited Count:3 Percentile:25.63(Chemistry, Analytical)

As one of methods for recovering uranium from the uranium-bearing wastes, we have proposed the electrolytic deposition method using choline chloride-urea (CCU) which is known as an ambient temperature molten salt. More than 92% of uranium components in inactivated alumina and spent sodium fluoride adsorbent was dissolved into CCU solution. Cyclic voltammograms (CVs) of the solutions prepared by dissolving uranium-bearing wastes in CCU were measured in the potential range of -2.0 to 1.1 V (vs. Ag/AgCl). The one reduction peak was observed around -0.7 V for all solutions. Based on the results of CVs, bulk electrolyses of the solutions dissolving uranium-bearing wastes were also carried out at -1.5V at 80 $$^{circ}$$C. The deposits were formed on a carbon electrode as cathode. Consequently, we confirmed that CCU is effective media for recovering uranium selectively from uranium-bearing waste.

Journal Articles

Feasibility studies on electrochemical recovery of uranium from solid wastes contaminated with uranium using 1-butyl-3-methylimidazorium chloride as an electrolyte

Ohashi, Yusuke; Harada, Masayuki*; Asanuma, Noriko*; Ikeda, Yasuhisa*

Journal of Nuclear Materials, 464, p.119 - 127, 2015/09

 Times Cited Count:11 Percentile:76.52(Materials Science, Multidisciplinary)

In order to examine feasibility of the electrochemical deposition method for recovering uranium from the solid wastes contaminated with uranium using ionic liquid as electrolyte, we have studied the electrochemical behavior of each solution prepared by soaking the spent NaF adsorbents and the steel waste contaminated with uranium in BMICl (1-butyl-3-methyl- imidazolium chloride). The uranyl(VI) species in BMICl solutions were found to be reduced to U(V) irreversibly around -0.8 to -1.3 V vs. Ag/AgCl. Based on the electrochemical data, we have performed potential controlled electrolysis of each solution at -1.5 V vs. Ag/AgCl. Black deposit was obtained, and their composition analyses suggest that the deposit is the mixtures of U(IV) and U(VI) compounds containing O, F, Cl, and N elements. From the present study, it is expected that the solid wastes contaminated with uranium can be decontaminated by treating them in BMICl and the dissolved uranium species are recovered electrolytically.

Journal Articles

Application of ionic liquid as a medium for treating waste contaminated with UF$$_{4}$$

Ohashi, Yusuke; Asanuma, Noriko*; Harada, Masayuki*; Wada, Yukio*; Matsubara, Tatsuo; Ikeda, Yasuhisa*

Journal of Nuclear Science and Technology, 46(8), p.771 - 775, 2009/08

Most of the metal or bed material wastes generated from uranium enrichment facilities or uranium refining and conversion plants are contaminated by uranium fluoride compounds such as UF$$_{4}$$. The UF$$_{4}$$ powder was completely dissolved in BMICl(1-buthyl-3-methylimidazolium chloride). The uranium concentrations of metal waste dropped below the temporary proposed clearance level (1.0 Bq/g) using BMICl. In the cyclic voltammogram of BMICl solution when dissolving UF$$_{4}$$, uncoupled reduction and oxidation peaks were observed and the reduction peak was considered to correspond to the reduction of uranyl(VI) + e$$^{-}$$ $$rightarrow$$ uranyl(V) followed by further reduction to UO$$_{2}$$.

Journal Articles

A Study on precipitation behavior of plutonium and other transuranium elements with N-cyclohexyl-2-pyrrolidone for development of a simple reprocessing process

Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Ikeda, Yasuhisa*; Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*

Journal of Nuclear Science and Technology, 44(3), p.354 - 360, 2007/03

 Times Cited Count:13 Percentile:69.08(Nuclear Science & Technology)

Precipitation behavior of Pu and other transuranium elements with N-cyclohexyl-2-pyrrolidone (NCP) has been examined to develop a simple reprocessing based only on precipitation method. From HNO$$_{3}$$ solutions containing only Pu, both Pu(VI) and Pu(IV) were precipitated with NCP, but they required more NCP than in the U(VI) precipitation. Selective U(VI) precipitation from HNO$$_{3}$$ solution containing U(VI) and Pu(IV) was achieved by stirring the solution for sufficient time after addition of NCP with ratio of [NCP]/[U]=1.4. Addition of an enough amount of NCP to U(VI)-Pu(VI) or U(VI)-Pu(IV) solutions gave a quantitative precipitation of both U and Pu. Neither Am(III) nor Np(V) was precipitated in the selective U precipitation and the simultaneous U-Pu precipitation. These results demonstrate the feasibility of the reprocessing by precipitation with NCP.

Journal Articles

Development of a simple reprocessing process using selective precipitant for uranyl ions; Precipitation behaviors of plutonium and other transuranium elements

Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Ikeda, Yasuhisa*; Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

N-cyclohexyl-2-pyrrolidone (NCP) can selectively precipitate U(VI) ions in aqueous nitric acid solutions. Utilizing this property, we have been developing a simple reprocessing process of spent nuclear fuel based only on precipitation method. In the first precipitation step, only U is separated by precipitation in a yield of about 70%, and in the second precipitation step both U and Pu are recovered and separated from fission products (FP) and other transuranium elements (TRU). In JAERI, precipitation behaviors of Pu and other TRU were examined experimentally, and the results showed the feasibility of the process establishement.

Journal Articles

Development of a simple reprocessing process using selective precipitant for uranyl ions; Engineering studies for precipitating and separating systems

Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*; Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Harada, Masayuki*; et al.

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 4 Pages, 2005/10

N-cyclohexyl-2-pyrrolidone (NCP), can selectively precipitate U(VI) ions in aqueous nitric acid solutions. Utilizing this property, we have been developing a simple reprocessing process of spent nuclear fuel based only on precipitation method. In the first precipitation step, only U is separated by precipitation in a yield of about 70%, and in the second precipitation step both U and Pu are recovered and separated from fission products (FP) and other transuranium elements (TRU). In the present study, a precipitator and a precipitate separator were designed and built up, and were tested with aspets of operationability and system performance.

Journal Articles

Tritium confinement demonstration using Caisson assembly for tritium safety study at TPL/JAERI

Hayashi, Takumi; Kobayashi, Kazuhiro; Iwai, Yasunori; Asanuma, Noriko; Ohira, Shigeru; Nishi, Masataka

Fusion Science and Technology, 41(3), p.647 - 651, 2002/05

no abstracts in English

Journal Articles

Tritium behavoir study for detritiation of atmosphere in a room

Kobayashi, Kazuhiro; Hayashi, Takumi; Iwai, Yasunori; Asanuma, Noriko; Nishi, Masataka

Fusion Science and Technology, 41(3), p.673 - 677, 2002/05

no abstracts in English

Oral presentation

Development of a simple reprocessing process by selective precipitation for uranyl ions

Kikuchi, Toshiaki*; Yamasaki, Kazuhiko*; Kusama, Makoto*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Hanzawa, Masatoshi*; Koshino, Nobuyoshi*; Asanuma, Noriko*; Harada, Masayuki*; Kawata, Yoshihisa*; et al.

no journal, , 

no abstracts in English

Oral presentation

Molybdenum extraction from spent nuclear fuel by using pyridine resin; Distribution coefficients of molybdenum and zirconium on pyridine resin

Suzuki, Tatsuya*; Honda, Masanori*; Asanuma, Noriko*; Osaka, Masahiko

no journal, , 

Pyridine-resin based ion-exchange method has been proposed as a recovery technique of molybdenum from high level waste. A batch experiment was carried out for the evaluation of Mo distribution coefficient to the pyridine-resin. Zirconium adsorption behavior was also investigated.

Oral presentation

Development of uranium recovery technique from materials contaminated with uranium using ionic liquid, 4; Electrolytic recovery of uranium from ionic liquid

Ohashi, Yusuke; Asanuma, Noriko*; Harada, Masayuki*; Ikeda, Yasuhisa*

no journal, , 

In the cyclic voltammogram of BMICl solution when dissolving metal waste contaminated with UF$$_{4}$$, uncoupled reduction peak and a pair of oxidation-reduction peaks were observed around -1.0V and 0V. The reduction peak was considered to correspond to the reduction of U(VI) $$rightarrow$$ U(V). A pair of peaks were considered to be the reversible peak of Fe(III) $$rightarrow$$ Fe(II). After the electrolysis of the solution, the working electrode was washed by ethanol. From the results of XPS measurement, uranium component was observed on the surface of the electrode though no Fe conponent was observed. Hence we would expect that the uranium component could be recovered by electrolysis from the solution generated in the decontamination treatment of the steel wastes contaminated with UF$$_{4}$$ in BMICl.

Oral presentation

Development of treatment processes of uranium contaminated wastes using adsorbent with high selectivity to uranium, 1; Dissolution behavior of sludge containing uranium into inorganic acid

Ohashi, Yusuke; Harada, Masayuki*; Asanuma, Noriko*; Ikeda, Yasuhisa*

no journal, , 

Solid and liquid radioactive waste derived from research and development for uranium refining and conversion has been stored at the Japan Atomic Energy Agency Ningyo-toge Environmental Engineering Center. The solid waste with uranium totals 1500ton.Most of this waste is considered to exceed the radioactivity level for VLLW. Hence we have proposed to recover uranium using adsorbent with high selectivity to uranium after dissolving sludge into inorganic acid. Dissolution experiments were carried out by adding each sludge sample into 1N hydrochloric acid as changing dissolution time. 90% of Uranium in calcium fluoride sludge dissolved within 10 minutes. Calcium fluoride itself didn't dissolved completely. Main component of active alumina adsorbent dissolved slowly though uranium in active alumina adsorbent dissolved completely within 10 minutes. Main component of sodium fluoride adsorbent and uranium in it dissolved completely within 10 minutes.

Oral presentation

Development of treatment processes of uranium contaminated wastes using adsorbent with high selectivity to uranium

Ohashi, Yusuke; Ando, Shion; Tsunashima, Yasumichi; Harada, Masayuki*; Ikeda, Yasuhisa*; Asanuma, Noriko*

no journal, , 

Sludge like uranium bearing waste and adsorbent were dissolved using hydrochloric acid. Uranium in the solution is recovered using PVPP. Recovered uranium was desorbed from PVPP using pure water and then uranium was recovered as uranium peroxide. As a result, it was confirmed that highly pure uranium was recovered.

Oral presentation

Electrochemical recovery of uranium from solid wastes contaminated with uranium using deep eutectic as electrolyte

Ohashi, Yusuke; Ikeda, Yasuhisa*; Harada, Masayuki*; Asanuma, Noriko*

no journal, , 

Approximately 60% of uranium in spent filter aids is found to be dissolved selectively in CCU at 100$$^{circ}$$C. An uncoupled reduction and some oxidation peaks were observed around -0.7 V and 0.4, 0.8 V, respectively. Potential controlled electrolysis of CCU dissolving the spent filter aids was carried out under the constant potential (-1.5 V). As a result, the black deposit were produced. The XPS analyses suggest that the deposits are the uranium compounds mainly containing O and F. We confirmed that CCU is effective media for recovering uranium selectively from spent filter aids by the electrolytic method.

Oral presentation

Basic study of separating ammonium ion by using zeolite adsorbent

Miyano, Riku*; Asanuma, Noriko*; Matsuura, Haruaki*; Aihara, Haruka; Watanabe, So; Nomura, Kazunori

no journal, , 

no abstracts in English

Oral presentation

Ammonium separation and decomposition for radioactive liquid waste treatment, 2; Separation of ammonium by zeolite absorbent as a pretreatment

Asanuma, Noriko*; Miyano, Riku*; Shimizu, Tomu*; Aihara, Haruka; Watanabe, So; Nomura, Kazunori

no journal, , 

no abstracts in English

Oral presentation

Development of treatment method for analytical waste solutions in STRAD project, 2; Ammonium ion adsorption onto zeolites

Asanuma, Noriko*; Miyano, Riku*; Aihara, Haruka; Watanabe, So; Nomura, Kazunori

no journal, , 

Oral presentation

Overview of STRAD project for systematic treatments of radioactive liquid wastes generated in nuclear facilities

Watanabe, So; Ogi, Hiromichi*; Arai, Yoichi; Aihara, Haruka; Shibata, Atsuhiro; Nomura, Kazunori; Kamiya, Yuichi*; Asanuma, Noriko*; Matsuura, Haruaki*; Kubota, Toshio*; et al.

no journal, , 

25 (Records 1-20 displayed on this page)