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Fukui, Toshiki*; Maki, Takashi*; Miura, Nobuyuki; Tsukada, Takeshi*
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 23(2), p.169 - 173, 2016/12
The basic research programs for the next generation vitrification technology, which are commissioned project from Ministry of Economy, Trade and Industry of Japan, have been implemented from 2014 until 2018 for developing the advanced vitrification technology of low level wastes and high level liquid wastes.
Morita, Yasuji; Yamagishi, Isao; Sato, Soichi; Kirishima, Akira*; Fujii, Toshiyuki*; Tsukada, Takeshi*; Kurosaki, Ken*
no journal, ,
Development of high-level liquid waste (HLW) conditioning technology for advanced nuclear fuel cycle was conducted for the purpose of the reduction of potential problems in the verification process for HLW. The present report introduces the outline of the research and development, which consists of Mo-Pd-Ru separation technology and advances treatment technology for dissolution residue.
Morita, Yasuji; Yamagishi, Isao; Sato, Soichi; Kirishima, Akira*; Fujii, Toshiyuki*; Tsukada, Takeshi*; Kurosaki, Ken*
no journal, ,
Development of high-level liquid waste (HLW) conditioning technology for advanced nuclear fuel cycle was conducted for the purpose of the reduction of potential problems in the verification process for HLW. The present report introduces the outline of the research and development, which consists of Mo-Pd-Ru separation technology and advances treatment technology for dissolution residue.
Usami, Tsuyoshi*; Tsukada, Takeshi*; Morita, Yasuji
no journal, ,
Development of high-level liquid waste (HLW) conditioning technology for advanced nuclear fuel cycle was conducted for the purpose of the reduction of potential problems in the verification process for HLW. It includes development of separation technology for Mo, Pd and Ru from HLW and development of separate treatment of the insoluble residue. To evaluate characteristics of the insoluble residue, simulated residue of metal alloy composed of Ru, Rh, Pd, Mo and Re was dissolved with heated nitric acid. The results showed that higher concentration of Pd and Mo in the alloy makes the alloy easier to be dissolved. The alloy without Pd was hardly dissolved by nitric acid. On the other hand, the alloy without Ru was dissolved easily.
Usami, Tsuyoshi*; Tsukada, Takeshi*; Yamagishi, Isao; Morita, Yasuji
no journal, ,
Development of high-level liquid waste (HLW) conditioning technology for advanced nuclear fuel cycle was conducted for the purpose of the reduction of potential problems in the verification process for HLW. It includes development of separation technology for Mo, Pd and Ru from HLW and development of separate treatment of the insoluble residue. To evaluate characteristics of the insoluble residue, simulated residue of metal alloy composed of Ru, Mo, Rh and Pd was dissolved with heated nitric acid. The results showed that higher concentration of Pd in the alloy makes the alloy easier to be dissolved, and higher concentration of Ru makes the alloy more difficult to be dissolved.
Morita, Yasuji; Yamagishi, Isao; Sato, Soichi; Kirishima, Akira*; Fujii, Toshiyuki*; Uehara, Akihiro*; Tsukada, Takeshi*; Usami, Tsuyoshi*; Kurosaki, Ken*
no journal, ,
Development of high-level liquid waste (HLW) conditioning technology for advanced nuclear fuel cycle was conducted for the purpose of the reduction of potential problems in the verification process for HLW. It includes development of separation technology for Mo, Pd and Ru from HLW and development of separate treatment of the insoluble residue. The present report gives the overall results of the research and development and their evaluation. For the Mo separation, the extraction process with HDEHP was developed by performing continuous extraction tests and process simulation by a calculation code. An extraction process for Pd by 5,8-diethyl-7-hydroxy-6-dodecanone oxime was also developed, but was evaluated as less mature than the HEDHP process. As Ru separation method, volatilization of RuO
after electrochemical oxidation was examined. Dissolution residue (metal alloy) and recovered Pd and Ru were solidified together by hot-press method.
Usami, Tsuyoshi*; Tsukada, Takeshi*; Morita, Yasuji
no journal, ,
Development of high-level liquid waste (HLW) conditioning technology for advanced nuclear fuel cycle was conducted for the purpose of the reduction of potential problems in the verification process for HLW. It includes development of separation technology for Mo, Pd and Ru from HLW and development of separate treatment of the insoluble residue. To evaluate characteristics of the insoluble residue, simulated residue of metal alloy composed of Ru, Rh, Pd, Mo and Re was dissolved with heated nitric acid. The results showed that higher concentration of Pd and Mo in the alloy makes the alloy easier to be dissolved. The alloy without Pd was hardly dissolved by nitric acid. On the other hand, the alloy without Ru was dissolved easily.
Uruga, Kazuyoshi*; Tsukada, Takeshi*; Yamagishi, Isao; Terada, Atsuhiko; Uchiyama, Hideaki*
no journal, ,
CIPPEI fabricated a small scale model of the spent zeolite adsorption vessel in Fukushima Nuclear Power Plant and performed to heating test at the center of the zeolite filling bed. As a result, the chloride concentration at the bottom of vessel decreases as time has passed. Chlorine concentrated around the adsorption vessel center.
Yamagishi, Isao; Kato, Chiaki; Nagaishi, Ryuji; Arisaka, Makoto; Uruga, Kazuyoshi*; Tsukada, Takeshi*
no journal, ,
no abstracts in English
Yoshioka, Masahiro*; Fukui, Toshiki*; Miura, Nobuyuki; Tsukada, Takeshi*
no journal, ,
The basic research programs for the next generation vitrification technology, which are commissioned project from Ministry of Economy, Trade and Industry of Japan to IHI Corporation (IHI), Japan Nuclear Fuel Limited (JNFL), Japan Atomic Energy Agency (JAEA) and Central Research Institute of Electric Power Industry (CRIEPI), have been implemented from 2014 for developing the advanced vitrification technology of low level wastes and high level liquid wastes (HLLW). In these programs, the developmental works such as the high waste loading glass, the alternate glasses of current borosilicate glasses including glass-ceramics and the minor actinide adsorbent glasses have been entrusted with the above organizations.
Ishio, Takahiro*; Kanehira, Norio*; Hoshino, Takeshi*; Fukui, Toshiki*; Iwabuchi, Hiroki; Tsukada, Takeshi*
no journal, ,
In Japan, the High Level radioactive Liquid Waste (HLLW) generated along with the nuclear fuel cycle is to be vitrified, and its vitrification technology has been made practicable. And, various kinds of Low Level radioactive Liquid Waste (LLW) generated from reprocessing plant and nuclear power plants in Japan have been primarily treated by various methods such as incineration, compaction, cement solidification, however, vitrification method have not been introduced. On the other hand, there is a potential generation of LLW which has relatively high radioactivity level in case of conducting the decommissioning of reprocessing plant and nuclear power plants. Therefore, various kinds of the solidification and the volume reduction technologies have been developed in order to ensure the stable forms with smaller volumes for the LLW disposal. Furthermore, if the foundation for LLW vitrification technology is developed, it can be reflected in the advancement of vitrification technology of HLLW. Therefore, the Ministry of Economy, Trade and Industry launched the project "Basic Research Programs of Vitrification Technology for Waste Volume Reduction" during FY 2014 - 2018. IHI Corporation (IHI), Japan Nuclear Fuel Limited (JNFL), Japan Atomic Energy Agency (JAEA) and Central Research Institute of Electric Power Industry (CRIEPI) have commissioned this project. The development goals for this project are as follows. (1) To develop LLW generated at nuclear power plants and reprocessing plant, etc., to reinforce the foundation of vitrification technology for high volume reduction and more stable waste. (2) To study also advanced improvement of vitrification of HLLW that is practically used in Japan, by reflecting the findings obtained from LLW infrastructures. In this presentation we will report on our past achievements and future plans in this project.
Uruga, Kazuyoshi*; Tsukada, Takeshi*; Terada, Atsuhiko; Yamagishi, Isao
no journal, ,
For the safe storage of zeolite wastes generated by the treatment of radioactive saline water at the Fukushima Daiichi Nuclear Power Station, this study investigated the fundamental transport properties of water and chloride in zeolite column.
Uruga, Kazuyoshi*; Tsukada, Takeshi*; Yamagishi, Isao; Terada, Atsuhiko
no journal, ,
The inside of the zeolite adsorption tower used for the treatment of contaminated water at the Fukushima Daiichi Nuclear Power Station is predicted during storage. As a result of constructing a model based on the test results and performing numerical analysis on the evaporation/condensation of the residual water inside and the permeation into the zeolite, it showed that the salt contained in the residual water was concentrated in the center of the adsorption vessel. Then, the salinity in the residual water decreased.
Sagawa, Yusuke*; Yamagishi, Isao; Terada, Atsuhiko; Uruga, Kazuyoshi*; Tsukada, Takeshi*
no journal, ,
no abstracts in English
Makino, Hitoshi; Asano, Hidekazu*; Usami, Tsuyoshi*; Tsukada, Takeshi*; Ikeda, Takao*; Kawai, Kota*; Watanabe, Daisuke*
no journal, ,
This presentation shows current status of discussion on "An assessment for MOX plu-thermal cycle" which is a challenge as part of an attempt to assess total performance of advanced nuclear fuel cycle in the Research Committee on Disposal of Radioactive Waste and Partitioning-Transmutation Technology.
Okamoto, Yoshihiro; Masuno, Atsunobu*; Owaku, Kohei*; Tsukada, Takeshi*; Kanehira, Norio*
no journal, ,
In the development of vitrification technology for high burnup fuel and MOX fuel, tests using various compositions of raw glass materials were conducted to find the optimum composition. The Si/B ratio and the amount of alkali in the raw glass, and even the loading ratio of waste components were varied. In this study, we summarize the composition dependence obtained by structural analysis of those samples.
