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Aoshima, Atsushi; Suzuki, Yoshiharu; Namekawa, Takashi
Nihon Genshiryoku Gakkai-Shi ATOMO, 55(12), p.733 - 736, 2013/12
no abstracts in English
Aoshima, Atsushi; Ueno, Tsutomu; Shiotsuki, Masao
Proceedings of 16th International Conference on Nuclear Engineering (ICONE-16) (CD-ROM), 9 Pages, 2008/05
In TVF, concentrated high radioactive liquid waste produced in TRP has been vitrified since 1995. Because of very corrosive condition of melting glass, design life of a melter is limited five years and it requires interruption of plant operation and generation of high radioactive solid waste for melter change. To improve this situation, prolongation of melter design life time by increasing corrosion resistance of structural material is required strongly. Effective removal of noble metal from a melter is also required because accumulated noble metal cause shortening lifetime of a melter. So, JAEA tried to develop a melter which equipped with ability of high corrosion resistance and control temperature distribution for noble metal easy drain out. Mechanical removal technology of remained noble metal rich glass also used if necessary. Low temperature glassing technology and advance removal of noble metal from concentrated high radioactive liquid are also be studying.
Aoshima, Atsushi; Ueno, Tsutomu; Shiotsuki, Masao
Proceedings of European Nuclear Conference 2007 (ENC 2007) (CD-ROM), 5 Pages, 2007/09
Tokai Vitrification Facility (TVF) started hot operation in 1996 and produced 241 canisters as of June 2007. Through TVF operation, JAEA had much experience and accumulated much technical know-how which indicated that management for noble metal accumulation in a melter was key technology for smooth plant operation. JAEA should continue service operation based on a vitrification contract with the Japanese utilities because there remains about two third of High Active Liquid Waste (HALW) produced in the reprocessing service operation of Tokai Reprocessing Plant (TRP). TVF melter is designed in condition of five years life time because of very corrosive characteristic of melted glass. Five years design life time is equivalent to 500 canisters production in TVF. Because estimated number of canisters which will be produced in the future is over 500 canisters, exchange of the present melter is necessary. From these situations, JAEA decided basic strategy to increase stability of the existing melter operation and develop an advanced new melter for replacement in future which has largely prolonged life time and high noble metal drain ability. To attain these targets, JAEA extracted necessary key technologies to assemble into a ten years road map and started development. This development has been progressing on schedule.
Shiotsuki, Masao; Aoshima, Atsushi; Nomura, Shigeo
Proceedings of International Waste Management Symposium 2006 (WM '06) (CD-ROM), 10 Pages, 2006/02
Achievement of reliable technologies on solidification and disposal of the HLW from future fuel cycle systems such for high burnup LWR, Pu-thermal (MOX), fast breeder reactor (FBR) and their transient stages is one of the most important issues to establish such advanced fuel cycle systems. In this paper, applicability and flexibility of the current vitrification technology for LWR fuel cycle to HLW from the future fuel cycle systems were reviewed by examining characteristics of the HLWs. The current developed vitrification technology is expected to have an advantage for applying to the solidification process of the HLW generated from future fuel cycle systems with some modification/optimization of the melting condition, etc. Moreover, it is thought that the advance aqueous reprocessing system developed for future FBR cycle has the potential which can contribute to the further reducing the number/volume of the HLW. It is also confirmed that development efforts on countermeasure for accumulation of noble metals, which JAEA has been carrying out aiming to accomplishing more stable and reliable operation of the vitrification process and extending the melter's life, will be able to contribute in the future fuel cycle system furthermore.
Aoshima, Atsushi; Tanaka, Kazuhiko
Proceedings of International Waste Management Symposium 2005 (WM '05) (CD-ROM), 10 Pages, 2005/03
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.
Aoshima, Atsushi; Tanaka, Kazuhiko ; Kozaka, Tetsuo
Proceedings of 12th International Conference on Nuclear Engineering (ICONE-12) (CD-ROM), 6 Pages, 2004/00
no abstracts in English
; ; Kamoshida, Mamoru*; Sasahira, Akira*
Journal of Nuclear Science and Technology, 39(Suppl.3), p.317 - 320, 2002/11
None
Morita, Yasuji; Tachimori, Shoichi; Koma, Yoshikazu*; Aoshima, Atsushi*
JAERI-Research 2002-017, 20 Pages, 2002/08
The present report describes the results of a joint study between Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Energy Research Institute (JAERI) on actinide separation process from high-level liquid waste. The purpose of the joint study is to point out common subjects in process development by an overall evaluation of each actinide separation process: TRUEX/SETFICS Process studied in JNC and DIDPA Extraction Process studied in JAERI. The result of the evaluation showed that both processes have common subjects to be studied in sub-processes such as treatment step for spent solvent or DTPA waste solution and solvent washing step for recycling, although the main process is different from each other. It is necessary to develop the sub-processes and to test the whole process including the sub-processes. Two essential requirements: the cost reduction and the minimization of secondary wastes, are very important in future research and development for more rational and effective actinide separation process.
; Oda, Yoshihiro;
Journal of Nuclear Science and Technology, 39(6), p.647 - 654, 2002/06
Times Cited Count:38 Percentile:89.69(Nuclear Science & Technology)None
*; *; *; *; *; ;
JNC TY8400 2002-009, 20 Pages, 2002/05
no abstracts in English
*; *; *; ;
JNC TY8400 2002-005, 42 Pages, 2002/05
no abstracts in English
*; *; *; *; Miyahara, Sachiko; ;
JNC TY8400 2002-004, 115 Pages, 2002/05
no abstracts in English
*; *; *; *; *; Nomura, Kazunori;
JNC TY8400 2002-003, 81 Pages, 2002/05
no abstracts in English
Miyahara, Sachiko; *; Shiba, Masanori*; *; ; *;
JNC TN8400 2002-014, 40 Pages, 2002/05
The current technology for the selective separation of plutonium and uranium from spent nuclear fuel (MOX) using TBP-HNO complex is being developed (Powdered fuel extraction process). It is promising to simplify the reprocessing process for the selective separation because of its potential to unite the chemical processes, dissolution process using nitric acid and co-extraction process using TBP solvent, and to operate under the ambient pressure and at relatively "mild" temperature. Plutonium oxide has reported to provide slower dissolution than uranium oxide in nitric acid. In this work dissolution behaviors of plutonium into TBP-HNO complex from powdered plutonium and uranium mixed oxide were examined. The powdered MOX fuel (average particles size 10m) was prepared from PuO-O pellets by heating for 4 hours at 400C. The prepared powder was dissolved into TBP-4.74mol/L HNO complex and was stirred for 300 minutes. In the test with 6 grams of powdered MOX fuel and 20 mL of the TBP-HNO complex, the concentration of plutonium reached 0.17 mol/L and about 90 percent of plutonium was dissolved. It is experimentally confirmed plutonium was dissolved into the TBP-HNO complex from plutonium and uranium mixed oxide. The early dissolution rate was almost the same as that obtained with nitric acid solution. It is likely to predict the dissolution rate from the rate for nitric acid solution. Americium that was contained in the MOX fuel was also dissolved into the TBP-HNO complex, but was slower than plutonium.
; ; Morita, Yasuji*; *
JNC TY8400 2002-001, 26 Pages, 2002/03
no abstracts in English
; ; Nomura, Kazunori;
Posters Session,Part, 0 Pages, 2002/00
None
; Miyachi, Shigehiko; ;
Proceedings of 10th International Conference on Nuclear Engineering (ICONE-10), 0 Pages, 2002/00
None
Nagai, Toshihisa; Ojima, Hisao; ;
Proceedings of 10th International Conference on Nuclear Engineering (ICONE-10), 0 Pages, 2002/00
None