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Namba, Takashi; Nagaoki, Yoshihiro; Sagayama, Yutaka
011-03, 0 Pages, 2004/06
Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.
Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Matsumura, Masakazu*; Kim, S.-Y.; Mineo, Hideaki; Morita, Yasuji
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles, 5 Pages, 2004/06
In JAERI, PARC process based on PUREX technique has been studied to as the basis of future reprocessing. The key of concept is to obtain the products, U and Pu, within only a single extraction cycle by separating Np and Tc from U and Pu before U/Pu partition. Two flow-sheet tests on the process were performed with 44 GWd/t PWR spent-fuel solutions. It was demonstrated that remaining Np in raffinate from co-extraction could be decreased to 13 % compared to the dissolver solution with increased solvent flow rate and with increased nitric acid concentration of FP scrubbing solution. It was demonstrated that Np separation (selective reduction by n-butyraldehyde) efficiency could be improved from 36 % to 78 % by flow-sheet modification; increasing reductant concentration and scrubbing solution flow rate. The feasibility of the Tc separation technique by high acid scrubbing was demonstrated.
Suzuki, Shinichi; Sasaki, Yuji; Yaita, Tsuyoshi; Kimura, Takaumi
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles (CD-ROM), 4 Pages, 2004/06
An innovative chemical separation process (ARTIST: Amide-based Radio-resources Treatment with Interim Storage of Transuranics) was proposed for the treatment of spent nuclear fuel. The main concept of ARTIST process is to recover and stock all actinides (An) and to dispose the fission products (FP). One of the main purposes of this process is selective isolation of uranium. Since the brached alkyl type N,N-dialkyl-monoamides (BAMA) have the steric hindrance on the complexation with metal cations, BAMA can be used to separate An(VI) from An(IV). N,N-di-(2-ethyl)hexyl-2,2-dimethylpropanamide (D2EHDMPA) can recover U(VI) selectively without accumulating Pu(IV) in uranium isolation process. From extraction behavior of Np, D2EHDMPA can extract and separate U(VI) from Np(VI) without reduction from Np(VI) to Np(V) or Np(IV).
Sasaki, Yuji; Sugo, Yumi; Suzuki, Hideya*; Kimura, Takaumi
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles (CD-ROM), 4 Pages, 2004/06
no abstracts in English
Arai, Yasuo; Pillon, S.*
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles (CD-ROM), 9 Pages, 2004/06
no abstracts in English
Ozawa, Masaki; *
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles (CD-ROM), 69 Pages, 2004/06
None
Complex for Super-DIREX ProcessKamiya, Masayoshi; Miura, Sachiko; Nomura, Kazunori; Koyama, Tomozo; Ogumo, Shinya*; Mori, Yukihide*; Enokida, Yoichi*
CD-ROM, P1-35, 4P., 4 Pages, 2004/00
Super-DIREX is a new reprocessing method which has high economical efficiency. Experimental study of this process was started on the direct extraction of U and Pu from irradiated MOX fuel by the supercritical carbon dioxide (SFCO
) containing TBP-HNO complex. This report describes direct extraction of U and Pu with TBP-HNO3 complex at atmospheric pressure, as the first test for irradiated fuel, in order to investigate the applicability of SFCO containing TBP-HNO complex. In this test, dependency on dissolution temperature, Pu content, fuel/ TBP-HNO complex ratio and effect of voloxidation were investigated. From these results, TBP-HNO complex was found to be effective in the respect of the recovery of U and Pu. The number of the process step in dissolution and co-extraction is small, and amount of waste can be reduced. It is applicable to the direct extraction in Super-DIREX.
Shibata, Atsuhiro; Yano, Kimihiko; Nomura, Kazunori; Koizumi, Tsutomu; Koyama, Tomozo; Miyake, Chie
CD-ROM, P1-66, 3P., 3 Pages, 2004/00
In the NEXT process, the greater part of uranium is separated from dissolver solution by crystallization. It is important to check the behavior of plutonium under the uranium crystallization operating condition in order to prevent plutonium from significantly accompanying uranium. In this study, Pu nitrate solution tests were performed.
Yano, Kimihiko; Shibata, Atsuhiro; Nomura, Kazunori; Koizumi, Tsutomu; Koyama, Tomozo; Miyake, Chie
CD-ROM, P1-66, 4p., 4 Pages, 2004/00
JNC has been developed uranium crystallization process as a component of advanced aqueous reprocessing system, NEXT (New Extraction System for TRU Recovery). In this process the greater part of uranium is separated from dissolver solution by crystallization as uranyl nitrate hexahydrate (UNH). Two types of experiments were carried out to clarify the plutonium behavior under the condition of uranium crystallization from dissolver solution. The results indicated that hexavalent plutonium is co-crystallized with uranium under the condition of uranium crystallization from dissolver solution, although plutonium concentration is lower than its solubility, and tetravalent plutonium is not. Therefore, the adjustment of plutonium valence at tetravalent is needed in order to avoid co-crystallization of plutonium and uranium.
Namba, Takashi; Funasaka, Hideyuki; Nagaoki, Yoshihiro; Sagayama, Yutaka
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles (CD-ROM), 0 Pages, 2004/00
Experimental Study on Temperature Fluctuation Phenomena at T-pipe Junction(6)