Current status and perspective of advanced loop type fast reactor in fast reactor cycle technology development project

Niwa, Hajime; Aoto, Kazumi; Morishita, Masaki

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.62 - 70, 2007/09

Conceptual design of the HTTR-IS hydrogen production system; Dynamic simulation code development for advanced process heat exchanger in the HTTR-IS system

Sato, Hiroyuki; Kubo, Shinji; Sakaba, Nariaki; Ohashi, Hirofumi; Sano, Naoki; Nishihara, Tetsuo; Kunitomi, Kazuhiko

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.812 - 819, 2007/09

The objective of this study is to confirm the availability of proposed mitigation methodology against thermal load increase events initiated by the thermochemical water splitting IS process hydrogen production system coupling with the HTTR. JAEA has been performing the development of dynamic simulation code which can evaluate complex phenomena in the HTTR-IS system all at one once to achieve the requirement. The notable feature of the developed code is the APHX module which enables to estimate the IS process thermal load variation considering phase change and chemical reaction behavior assumed in the APHX. In this paper, two cases of dynamic calculation for the thermal load increase events were performed using the newly developed APHX module. The results of the analytical studies clearly show the availability of the developed model for dynamic simulation of the HTTR-IS system and the thermal load increase mitigation methodology.

Attempts to improve radiolytic stability of amidic extractants

Sugo, Yumi; Sasaki, Yuji; Kimura, Takaumi; Sekine, Tsutomu*

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1870 - 1873, 2007/09

A tridentate diamide, -tetraoctyldiglycolamide (TODGA) is very useful for the recovery of actinide ions from spent nuclear fuel. Based on the mechanism of the radiolysis of TODGA in organic solution, an improvement of radiolytic stability of amidic extractants was attempted. The radiolytic degradation of TODGA was suppressed by the addition of appropriate compounds, due to reduction in the mole fraction of -dodecane. In addition, by using the solvents having low ionization potentials, TODGA could be protected from radiation. Because the charge transfer reaction in the primary process was inhibited. It was also confirmed that aromatic substituents in the molecule effectively improved the radiolytic stability.

Calculation of the pressure vessel failure fraction of fuel particle of gas turbine high temperature reactor 300C

Aihara, Jun; Ueta, Shohei; Mozumi, Yasuhiro; Sato, Hiroyuki; Motohashi, Yoshinobu*; Sawa, Kazuhiro

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.416 - 422, 2007/09

In high temperature gas-cooled reactors (HTGRs), coated particles are used as fuels. For upgrading HTGR technologies, present SiC coating layer which is used as the 3rd layer could be replaced with ZrC coating layer which have much higher temperature stability in addition to higher resistance to chemical attack by fission product palladium than the SiC coating layer. The ZrC layer could deform plastically at high temperatures. Therefore, the Japan Atomic Energy Agency modified an existing pressure vessel failure fraction calculation code to treat the plastic deformation of the 3rd layer in order to predict failure fraction of ZrC coated particle under irradiation. Finite element method is employed to calculate the stress in each coating layer. The pressure vessel failure fraction of the coated fuel particles under normal operating condition of GTHTR300C is calculated by the modified code. The failure fraction is evaluated as low as 3.510.

Waste handling activities in glovebox dismantling facility

Kitamura, Akihiro; Okada, Takashi; Kashiro, Kashio; Yoshino, Masanori*; Hirano, Hiroshi*

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.531 - 536, 2007/09

We present our waste handling activities in glovebox dismantling facility, installed in Plutonium Fuel Production Facility, Nuclear Fuel Cycle Engineering Laboratories, JAEA. In this facility, we treat only one size gloveboxes (3m3m1m), but for the future waste treatment, we segregate waste into material categories. We analyzed the data collected for the future decommissioning, waste treatment and waste disposal. We also present the improvements which are already made and will be made in the near future.

Launch of fast reactor cycle technology development project in Japan

Sagayama, Yutaka

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.251 - 258, 2007/09

JAEA launched a new FR Cycle Technology Development (FaCT) Project in cooperation with the Japanese electric utilities. The FaCT project is based on the conclusion of the Feasibility Study on Commercialized FR Cycle Systems (FS), which carried out in last seven years. In the FS, the combination of the sodium-cooled FR with oxide fuel, the advanced aqueous reprocessing and the simplified pelletizing fuel fabrication was selected as the main concept which should be developed principally. A conceptual design study of the main concept and R&D of innovative technologies are implemented toward an important milestone at 2015. The development targets, which were set up at the beginning stage of FS, were revised for the FaCT project based on the results of FS and change in Japanese society environment and in the world situation. International collaboration is promoted to pursue fast reactor cycle technology which deserves the global standard and its efficient development.

Development of advanced reprocessing system based on use of pyrrolidone derivatives as novel precipitants with high selectivity and control ability, 1; Concept of advanced reprocessing system and precipitation behavior of U(VI)

Ikeda, Yasuhisa*; Takao, Koichiro*; Harada, Masayuki*; Morita, Yasuji; Nogami, Masanobu*; Nishimura, Kenji*

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1503 - 1507, 2007/09

We have developed a reprocessing process for spent FBR fuels based on the precipitation method using pyrrolidone derivatives. In previous investigation, N-cyclohexyl-2-pyrrolidone (NCP) is used as a precipitant and a process consisting of selective U precipitation step and U-Pu co-precipitation step was developed. In the present study, in order to examine the applicability of precipitants with lower hydrophobicity than NCP to the selective U precipitation step, we have carried out precipitation experiments of U(VI) by N-butyl-2-pyrrolidone (NBP) and N-propyl-2-pyrrolidone (NProP) and measured decontamination factors of some fission products.

Progress in the R&D project on oxide dispersion strengthened and precipitation hardened ferritic steels for sodium cooled fast breeder reactor fuels

Kaito, Takeji; Otsuka, Satoshi; Inoue, Masaki

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.37 - 42, 2007/09

High burnup capability of sodium cooled fast breeder reactor (SFR) fuels depends significantly on irradiation performance of their component materials. Japan Atomic Energy Agency (JAEA) has been developing oxide dispersion strengthened (ODS) ferritic steels and a precipitation hardened (PH) ferritic steel as the most prospective candidate materials for fuel pin cladding and duct tubes, respectively. Technology for small-scale manufacturing is already established, and several hundreds of ODS steel cladding tubes and dozens of PH steel duct tubes were successfully produced. We will step forward to develop manufacturing technology for mass production to supply these steels for future commercialized SFRs. Mechanical properties of the products were examined by out-of-pile and in-pile tests including material irradiation tests in the experimental fast reactor JOYO and the Fast Flux Test Facility (FFTF). The material strength standards (MSSs) were tentatively compiled in 2005 for ODS steels and in 1993 for PH steel. In order to improve the MSSs and to demonstrate high burnup capability of the materials, we will perform a series of irradiation tests in BOR-60 and JOYO until 2015 and contribute to design study for a demonstration SFR of which operation is expected after 2025.

Study on characteristics of recycled MOX powder suitable for low density pellet fabrication

Murakami, Tatsutoshi; Suzuki, Kiichi; Aono, Shigenori

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.891 - 896, 2007/09

no abstracts in English

Separation of Actinide(III) from Lanthanide(III) by thermosensitive gel copolymerized with TPEN derivatives

Takeshita, Kenji*; Fugate, G.*; Matsumura, Tatsuro; Nakano, Yoshio*; Mori, Atsunori*; Fukuoka, Sachio*

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.120 - 125, 2007/09

Extraction separation of Am(III) and Eu(III) was examined by the thermal-swing extraction technique using a thermosensitive gel, poly--isopropylacrylamide (NIPA) copolymerized with a TPEN derivative, -tetrakis(4-propenyloxy-2-pyridylmethyl)ethylenediamine (TPPEN). The separation of Am(III) from Eu(III) was observed in the swollen state of gel (5C) and the separation factor of Am(III) was evaluated as about 18 at pH5.2. More than 90% of Am(III) extracted into the gel was released by the volume phase transition of gel from the swollen state (5C) to the shrunken one (40C). The repetition test for the thermal swing extraction of a soft metal ion, Cd(II), which was used as a substitute of Am(III), was carried out and the extraction and release of Cd(II) were repeated three times stably under the thermal-swing operation between 5C and 40C. The radiation effect of gel on the extraction of Am and Eu was tested by the irradiation of -ray (10 kGy) and the long-term adsorption of -emitter (Cm). The TPPEN-NIPA gel sustained no damage by these radiation tests. These results suggest that the thermal-swing extraction technique is applicable to the MA partitioning process indispensable for the establishment of P&T technology.