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Journal Articles

Study of treatment scenarios for fuel debris removed from Fukushima Daiichi NPS

Washiya, Tadahiro; Yano, Kimihiko; Kaji, Naoya; Yamada, Seiya*; Kamiya, Masayoshi

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

On March 11, 2011, a severe nuclear accident occurred at Tokyo Electric Power Company (TEPCO)'s Fukushima Daiichi Nuclear Power Plant (hereinafter called as F1). After the accident, the Council for the Decommissioning was established, mainly by the government and TEPCO, and a road map for the F1 decommissioning was drawn up. In the road map, the fuel debris removal from the reactors is scheduled to launch around 2020. In this study, the characteristics and technological issues of each potential treatment scenario were extracted, and the scenarios were prioritized in advance of formal evaluations in the future. The preliminary evaluation results show that long term storage and direct disposal have more positive aspects in terms of economic efficiency and radioactive waste generation. On the other hand, stabilizing processing, aqueous processing, and pyrochemical processing have been estimated to have more disadvantages in such aspects.

Journal Articles

Dissolution behavior of (U,Zr)O$$_{2}$$-based simulated fuel debris in nitric acid

Ikeuchi, Hirotomo; Ishihara, Miho; Yano, Kimihiko; Kaji, Naoya; Nakajima, Yasuo; Washiya, Tadahiro

Journal of Nuclear Science and Technology, 51(7-8), p.996 - 1005, 2014/07

 Times Cited Count:6 Percentile:50.1(Nuclear Science & Technology)

Journal Articles

Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi Nuclear Power Station

Ikeuchi, Hirotomo; Kondo, Yoshikazu*; Noguchi, Yoshihiro*; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1349 - 1356, 2013/09

Journal Articles

Direction on characterization of fuel debris for defueling process in Fukushima Daiichi Nuclear Power Station

Yano, Kimihiko; Kitagaki, Toru; Ikeuchi, Hirotomo; Wakui, Ryohei; Higuchi, Hidetoshi; Kaji, Naoya; Koizumi, Kenji; Washiya, Tadahiro

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1554 - 1559, 2013/09

Journal Articles

Nitric acid concentration dependence of dicesium plutonium(IV) nitrate formation during solution growth of uranyl nitrate hexahydrate

Nakahara, Masaumi; Kaji, Naoya; Yano, Kimihiko; Shibata, Atsuhiro; Takeuchi, Masayuki; Okano, Masanori; Kuno, Takehiko

Journal of Chemical Engineering of Japan, 46(1), p.56 - 62, 2013/01

 Times Cited Count:1 Percentile:7.24(Engineering, Chemical)

The influence of HNO$$_{3}$$ concentration in the solution on the formation of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ was evaluated in the U crystallization process. The solubility of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in a uranyl nitrate solution was found to decrease with increasing HNO$$_{3}$$ concentration in the solution. In the U crystallization experiments with the dissolver solution of irradiated fast reactor fuel, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ formed with 6.5 mol/dm$$^{3}$$ HNO$$_{3}$$ concentration in the mother liquor, and the decontamination factor of Cs for the uranyl nitrate hexahydrate crystals was low. Meanwhile, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ did not precipitate with uranyl nitrate hexahydrate crystals under the condition of 4.0 mol/dm$$^{3}$$ HNO$$_{3}$$ concentration in the mother liquor, and Cs could be separated from the uranyl nitrate hexahydrate crystals.

JAEA Reports

Elution properties of cesium contained within irradiated fast neutron reactor fuel in water and diluted nitric acid solution

Nakahara, Masaumi; Kaji, Naoya; Nomura, Kazunori

JAEA-Research 2012-009, 15 Pages, 2012/06

JAEA-Research-2012-009.pdf:6.37MB

In terms of preventing the formation of Pu and Cs compound, Cs in the feed solution should decrease in the U crystallization process. In order to separate Cs contained within irradiated nuclear fuel, the immersion experiments were carried out with the pure water and diluted HNO$$_{3}$$ solution. The elusion ratio of Cs within the powdered fuel in the pure water and 0.1 mol/dm$$^{3}$$ HNO$$_{3}$$ solution after 67 h was 33.8 and 38.3%, respectively. The experimental results suggest a possible beneficial effect of Cs elusion by immersion of the powdered fuel in the pure water and diluted HNO$$_{3}$$ solution before the fuel dissolution process.

Journal Articles

Salt-free technique for solvent washing process in NEXT process

Sano, Yuichi; Kaji, Naoya; Shibata, Atsuhiro; Takeuchi, Masayuki; Washiya, Tadahiro

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12

Journal Articles

A Consideration on proliferation resistance of a FBR fuel cycle system

Inoue, Naoko; Kaji, Naoya; Suda, Kazunori; Kawakubo, Yoko; Suzuki, Mitsutoshi; Koyama, Tomozo; Kuno, Yusuke; Senzaki, Masao

Proceedings of INMM 51st Annual Meeting (CD-ROM), 10 Pages, 2010/07

Journal Articles

Current status on research and development of uranium crystallization system in advanced aqueous reprocessing of FaCT project

Shibata, Atsuhiro; Kaji, Naoya; Nakahara, Masaumi; Yano, Kimihiko; Tayama, Toshimitsu; Nakamura, Kazuhito; Washiya, Tadahiro; Myochin, Munetaka; Chikazawa, Takahiro*; Kikuchi, Toshiaki*

Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.151 - 157, 2009/09

As a part of FaCT project, Japan Atomic Energy Agency has been developing a U crystallization process for advanced aqueous reprocessing technology in collaboration with Mitsubishi Materials Corporation. We have carried out experimental studies and obtained fundamental data. Continuous operation tests were also carried out by an engineering-scale crystallizer to confirm productivity of the equipment and to investigate non-steady state conditions. The requirements for the U crystallization process in the FaCT project could be achieved except DF of Cs. More detail investigation is under way to settle the process condition without Pu-Cs double salt formation.

Journal Articles

Dissolution of powdered spent fuel and U crystallization from actual dissolver solution for "NEXT" process development

Nomura, Kazunori; Hinai, Hiroshi; Nakahara, Masaumi; Kaji, Naoya; Kamiya, Masayoshi; Oyama, Koichi; Sano, Yuichi; Washiya, Tadahiro; Komaki, Jun

Proceedings of 3rd International ATALANTE Conference (ATALANTE 2008) (CD-ROM), 5 Pages, 2008/05

Oral presentation

Fundamental test of crystallization process; Evaluation of U solubility in solution of irradiated fuel

Kaji, Naoya; Nakahara, Masaumi; Nakamura, Kazuhito; Shibata, Atsuhiro; Tomita, Yutaka; Washiya, Tadahiro; Kitajima, Takafumi; Koizumi, Tsutomu

no journal, , 

Solubility obtained from the latest crystallization tests using irradiated fuel and the tests implemented before using U or Pu/U are compared with the data showed by Hart. Based on the result, availability of the data to estimate crystallization ratio is considered.

Oral presentation

Development of salt-free technology for FBR fuel reprocessing

Kaji, Naoya; Nomura, Kazunori; Sano, Yuichi; Koizumi, Tsutomu

no journal, , 

Advanced aqueous reprocessing is the technology that JAEA is developing as a part of Fast Reactor Cycle Technology Development Project (FaCT). Radioactive liquid waste polarization is one of the concepts introduced to the advanced aqueous reprocessing. In this concept, waste solutions are to be concentrated enough to just be divided to two categories of high level and very low level. It will be achieved by decreasing amount of sodium salt waste drastically and decomposing excess nitric acid. One of the main sources of sodium salt waste in the conventional PUREX is the solvent washing process with sodium carbonate. Salt-free technology brings alternative solvent washing method that does not produce the sodium salt waste. The outline and the current status of the development of this technology will be reported.

Oral presentation

Fundamental test of crystallization process; Experience on behavior of FPs in crystallization process by using irradiated MOX fuel

Kaji, Naoya; Nakamura, Kazuhito; Nomura, Kazunori; Shibata, Atsuhiro; Yano, Kimihiko; Washiya, Tadahiro; Kitajima, Takafumi; Koizumi, Tsutomu

no journal, , 

For the purpose of obtaining DFs of the short-life FPs by $$gamma$$ analysis in the hot experiment, fundamental test of crystallization process was carried out using Fast Experimental Reactor "JOYO" irradiated fuel. In this experiment, DFs of nuclides analyzed by $$gamma$$ analysis except Cs (Such as Ru-106, Ce-144, Pr-144) and Pu were evaluated to be about 100. It was suggested that there is the condition in which Pu-Cs double salt does not be formed, although DF of Cs had been very low in the past experiments because of forming Pu-Cs double salt.

Oral presentation

The Examination for applying salt-free solvent washing reagent to Tokai Reprocessing Plant

Yamamoto, Kohei; Kaji, Naoya; Morimoto, Kazuyuki; Obu, Tomoyuki; Sano, Yuichi; Kashimura, Takao

no journal, , 

no abstracts in English

Oral presentation

Development of actinides co-extraction system with direct extraction process using supercritical fluid, 10; Direct extraction test of unirradiated MOX fuel under supercritical condition

Kaji, Naoya; Kamiya, Masayoshi; Koma, Yoshikazu; Koizumi, Tsutomu; Koyama, Tomozo; Aoki, Kazuo*; Yamada, Seiya*

no journal, , 

no abstracts in English

Oral presentation

Development of actinides co-extraction system with direct extraction process using supercritical fluid, 12; The Outline of the other results of development

Kamiya, Masayoshi; Kaji, Naoya; Koyama, Tomozo; Aoki, Kazuo*; Yamada, Seiya*

no journal, , 

It is shown in this presentation that the outline "development of actinides co-extraction system with direct extraction process using supercritical fluid" and future schedule.

Oral presentation

Proliferation resistance of FBR cycle

Kaji, Naoya

no journal, , 

no abstracts in English

Oral presentation

Development of actinides co-extraction system with direct extraction process using supercritical fruid, 14; Direct extraction test of unirradiated MOX fuel

Kaji, Naoya; Kamiya, Masayoshi; Takahatake, Yoko; Oyama, Koichi; Miura, Sachiko; Koyama, Tomozo; Aoki, Kazuo*; Yamada, Seiya*

no journal, , 

A battery of tests using MOX fuel were performed for development of actinides co-extraction system with direct extraction process using supercritical fluid. The condition that was set up based on consideration of the results of the tests under ordinary pressure, and then was made fine adjusted based on the tests with supercritical fruid. As a result, U, Pu and Am were extracted simultaneously. Comparative test confirmed that U, Pu and Am could be also extracted simultaneously with dodecane as a diluent. While the extraction test of unirradiated MOX fuel with simulated FP, U, Pu was extracted with U and Am, and the Pu residue, that was found at the extraction test of irradiated fuel, was not found. As the result of scrubbing and stripping test, it was showed that traditional extraction process design method could be applied to design scrubbing and stripping process with supercritical fluid.

Oral presentation

Development of actinides co-extraction system with direct extraction process using supercritical fluid, 13; Scope

Koyama, Tomozo; Kamiya, Masayoshi; Kaji, Naoya; Aoki, Kazuo*; Yamada, Seiya*

no journal, , 

The summary of development of actinides co-extraction system with direct extraction process using supercritical fluid which was started at 2005 is reported here.

Oral presentation

Development of actinides co-extraction system with direct extraction process using supercritical fluid, 16; Integrated evaluation

Kamiya, Masayoshi; Kaji, Naoya; Koyama, Tomozo; Aoki, Kazuo*; Yamada, Seiya*

no journal, , 

Main process of actinides co-extraction system with direct extraction process using supercritical fluid was designed. The improvement effect of the cost performance was small because multiple lines were needed about main process. But we got the evaluation result that it is possible to reduce substantially about waste liquid treatment. Based on the result, the improvement which uses supercritical fluid for partitioning was gathered.

36 (Records 1-20 displayed on this page)