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Journal Articles

Development of nitride fuel cycle technology for transmutation of minor actinides

Hayashi, Hirokazu; Nishi, Tsuyoshi*; Sato, Takumi; Kurata, Masaki

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1811 - 1817, 2015/09

Transmutation of long-lived radioactive nuclides including minor actinides (MA: Np, Am, Cm) has been studied in Japan Atomic Energy Agency (JAEA). Accelerator-driven system (ADS) is regarded as one of the powerful tools for transmutation of MA under the double strata fuel cycle concept. Uranium-free nitride fuel was chosen as the first candidate fuel for MA transmutation using ADS. To improve the transmutation ratio of MA, reprocessing of spent fuel and reusing MA recovered from the spent fuels is necessary. Our target is to transmute 99% of MA arisen from commercial power reactor fuel cycle, with which the period until the radiotoxicity drops below that of natural uranium can be shorten from about 5000 years to about 300 years. A pyrochemical process has been proposed as the first candidate for reprocessing of the spent nitride fuel. This paper overviews the current status of the nitride fuel cycle technology. Our recent study on fuel fabrication, fuel property measurements, reprocessing of spent fuel, development of the property database of MA nitride fuel, and fuel behavior simulation code are introduced. Our research and development (R&D) plan based on the roadmap of the development is also introduced.

Journal Articles

Experiments on the behavior of americium in pyrochemical process

Hayashi, Hirokazu; Akabori, Mitsuo; Minato, Kazuo

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 3 Pages, 2005/10

For a basis of the future nuclear cycle, it is very important to understand and control the behavior of TRU (Np, Pu, Am, Cm) in the nuclear fuel cycle. Experimental study of pyrochemical process of fuels containing TRU requires the facility having not only shielding for $$gamma$$-ray and neutron but also ability to keep a high purity inert gas atmosphere; because minor actinide chlorides can easily react with oxygen or water vapor in an atmosphere. The module for TRU high temperature chemistry (TRU-HITEC) had been installed to study the basic properties of TRU in the pyrochemical processes. In the present work, the behavior of $$^{241}$$Am in pyrochemical process was investigated by electrochemical methods.

JAEA Reports

Proceedings of 4th Workshop on Molten Salts Technology and Computer Simulation; December 20, 2004, JAERI, Tokai, Japan

Research Group for Actinides Science

JAERI-Conf 2005-008, 216 Pages, 2005/09


This report is the Proceedings of the 4th Workshop on Molten Salts Technology and Computer Simulation, which was held on December 20, 2004, at Tokai Research Establishment of Japan Atomic Energy Research Institute (JAERI). The purpose of this workshop is to exchange information and views on molten salts technology and computer simulation among the specialists from domestic organizations, and to discuss the recent and future research status for this research field. The intensive discussion was made among approximately 55 participants. The presentations were 14 papers including one keynote lecture.

Journal Articles

Fabrication and electrochemical behavior of nitride fuel for future applications

Arai, Yasuo; Minato, Kazuo

Journal of Nuclear Materials, 344(1-3), p.180 - 185, 2005/09

 Times Cited Count:22 Percentile:82.13(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Simulation of chemical and electrochemical behavior of actinides and fission products in pyrochemical reprocessing

Minato, Kazuo; Hayashi, Hirokazu; Mizuguchi, Koji*; Sato, Takeyuki*; Amano, Osamu*; Miyamoto, Satoshi*

Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.778 - 781, 2003/11

The simulation technology for the pyrochemical reprocessing of oxide fuels was developed to analyze experimental data, to predict experimental results, and to propose adequate conditions and processes. The simulation method was based on calculations of chemical equilibrium and electrochemical reactions. Some model calculations to simulate the experimental results were made on the process of electro-codeposition of UO$$_{2}$$ and PuO$$_{2}$$. Although it was difficult to trace the experiments and compare the calculated results with the experimental results quantitatively due to the limitation of available data on the experimental conditions, the calculated results were consistent with the experimental results. The phenomena of the repeated oxidation-reduction reactions between Pu$$^{4+}$$ and Pu$$^{3+}$$ ions and those between Fe$$^{3+}$$ and Fe$$^{2+}$$ ions were theoretically analyzed,which caused the low current efficiency in the electro-codeposition process.

Journal Articles

Research and development on nuclear transmutation, B; Transmutation fuel and reprocessing

Minato, Kazuo; Arai, Yasuo

Genshikaku Kenkyu, 47(6), p.31 - 38, 2003/06

no abstracts in English

Journal Articles

Research and development on accelerator-driven system for transmutation of long-lived nuclear waste at JAERI

Oigawa, Hiroyuki; Sasa, Toshinobu; Takano, Hideki; Tsujimoto, Kazufumi; Nishihara, Kenji; Kikuchi, Kenji; Kurata, Yuji; Saito, Shigeru; Futakawa, Masatoshi; Umeno, Makoto*; et al.

Proceedings of 13th Pacific Basin Nuclear Conference (PBNC 2002) (CD-ROM), 8 Pages, 2002/10

To reduce the burden on the final disposal of the nuclear waste, the Acclelerator-Driven System (ADS) which can transmute minor actinides efficiently has been studied in JAERI. The proposed ADS design is an 800MWth subcritical core with lead-bismuth coolant and minor-actinide nitride fuel driven by a neutron source of a superconductivity linear accelerator with 30MW and a lead-bismuth spallation target. To realize the ADS, many research and development are under way in the fields of the accelerator, the spallation target and the nitride fuel. Moreover, a new experimental facility, the Transmutation Experimental Facility, is proposed under a framework of the High-Intensity Proton Accelerator Project to study the feasibility of the ADS from physics and engineering aspects.

JAEA Reports

Proceedings of the Workshop on Molten Salts Technology and Computer Simulation

Hayashi, Hirokazu; Minato, Kazuo

JAERI-Conf 2001-016, 181 Pages, 2001/12


Applications of molten salts technology to separations and syntheses of materials have been studied eagerly, which would develop new fields of materials science. Research Group for Actinides Science, Department of Materials Science, Japan Atomic Energy Research Institute (JAERI), together with Reprocessing and Recycle Technology Division, Atomic Energy Society of Japan, organized the Workshop on Molten Salts Technology and Computer Simulation at Tokai Research Establishment, JAERI on September 18, 2001. In the workshop eleven lectures were made and lively discussions were there on the bases and applications of the molten salts technology that covered the structure and basic properties of molten salts, the pyrochemical reprocessing technology and the relevant computer simulation.

Journal Articles

Thermochemical properties of advanced fission fuel materials

Ogawa, Toru; Okamoto, Yoshihiro; R.J.K.Konings*

Advances in Science and Technology, 24, p.381 - 392, 1999/00

no abstracts in English

Journal Articles

Dense fuel cycles for actinide burning and thermodynamic database

Ogawa, Toru; ; Kobayashi, Fumiaki; Ito, Akinori; Mukaiyama, Takehiko; Handa, Nuneo; R.G.Haire*

Global 1995,Int. Conf. on Evaluation of Emerging Nuclear Fuel Cycle Systems, 1, p.207 - 214, 1995/00

no abstracts in English

Oral presentation

Fuel cycle for MA transmutation based on nitride fuel and pyrochemical reprocessing

Takano, Masahide

no journal, , 

Concept and R&D status of the fuel cycle for MA transmutation based on nitride fuel and pyrochemical reprocessing are introduced in this lecture. Outline of our achievements in fabrication technologies for MA-bearing nitride fuel, thermal properties and establishment of the database, development of the nitride fuel performance code, issues of helium accumulation, and enrichment of nitrogen 15 is shown with the future plan for irradiation test and post-irradiation examination.

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