Characterization of the insoluble sludge from the dissolution of irradiated fast breeder reactor fuel

Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, Kazunori; Takeuchi, Masayuki

Procedia Chemistry, 21, p.279 - 284, 2016/12

BB2015-3214.pdf:0.31MB

https://doi.org/10.1016/j.proche.2016.10.039

Simulation study of sludge precipitation in spent fuel reprocessing

Takeuchi, Masayuki; Aihara, Haruka; Nakahara, Masaumi; Tanaka, Kotaro*

Procedia Chemistry, 21, p.182 - 189, 2016/12

BB2016-0225.pdf:0.61MB

https://doi.org/10.1016/j.proche.2016.10.026

A simulation technology with electrolyte thermodynamic model has been developed to evaluate the precipitation behavior in reprocessing solution based on nitric acid solution. The simulation results were compared with the experiment data from non-radioactive simulated HLLW with ten elements and Pu-Zr-Mo solution, and the reliability of the thermodynamic model was verified. Most of the precipitation species was zirconium molybdate hydrate from the both data. It is demonstrated that the chemical species and amount of the precipitation calculated by thermodynamic model reflected well that of experiments. This study has shown the thermodynamic simulation model is one of the useful tools to estimate the behavior of precipitation from the reprocessing solution.

Uranium and plutonium extraction by -dialkylamides using multistage mixer-settler extractors

Ban, Yasutoshi; Hotoku, Shinobu; Tsutsui, Nao; Suzuki, Asuka; Tsubata, Yasuhiro; Matsumura, Tatsuro

Procedia Chemistry, 21, p.156 - 161, 2016/12

https://doi.org/10.1016/j.proche.2016.10.022

A continuous counter-current experiment was carried out to demonstrate the validity of a process using -dialkylamides for recovering U and Pu. This process consisted of two cycles, and the 1st cycle and the 2nd cycle employed -di(2-ethylhexyl)-2,2-dimethylpropanamide and -di(2-ethylhexyl)butanamide as extractants, respectively. The feed solution for the 1st cycle was 5.1 mol/dm (M) nitric acid containing 0.92 M U, 1.6 mM Pu, and 0.6 mM Np. The raffinate collected in the 1st cycle was used as the feed for the 2nd cycle. The ratios of U recovered in the U fraction and U-Pu fraction were 99.1% and 0.8%, respectively. The ratio of Pu recovered in the U-Pu fraction was 99.7%. The concentration ratio of U with respect to Pu in the U-Pu fraction was 9, and this indicated that Pu was not isolated. The decontamination factor of U with respect to Pu in the U fraction was obtained as 4.510. These results supported the validity of the proposed process.

Dissolution behavior of irradiated mixed-oxide fuels with different plutonium contents

Ikeuchi, Hirotomo; Shibata, Atsuhiro; Sano, Yuichi; Koizumi, Tsutomu

Procedia Chemistry, 7, p.77 - 83, 2012/00

https://doi.org/10.1016/j.proche.2012.10.014

The effects of Pu content were studied on the dissolution rate of irradiated mixed oxide fuel and on the mass of insoluble residue. Kinetic analysis was conducted being based on the surface-reaction model to estimate the dissolution rate of irradiated fuels with Pu contents less than 30% and with burn-up ranging from 40.1 - 63.7 GWD/t. The dissolution rate of irradiated mixed-oxide fuels was found to decrease exponentially with an increase of the Pu content, but those were estimated to be up to 1000 times larger than those of non-irradiated fuels with the same Pu content. The amount of insoluble residue was found to increase with increase of the Pu content, possibly due to the promotion of fission product formation. Up to 1.3% of initial heavy metal was remained as the residue.

Advanced-ORIENT cycle project; Summary of phase I fundamental studies

Koyama, Shinichi; Suzuki, Tatsuya*; Ozawa, Masaki*; Kurosawa, Kiyoko*; Fujita, Reiko*; Mimura, Hitoshi*; Okada, Ken*; Morita, Yasuji; Fujii, Yasuhiko*

Procedia Chemistry, 7, p.222 - 230, 2012/00

https://doi.org/10.1016/j.proche.2012.10.037

Adv.-ORIENT cycle strategy has been proposed as a basic concept for trinitarian research on separation, transmutation and utilization of nuclides and elements based on FBR fuel cycle. Validation of principal separation method and related safety research were performed from 2006 through 2011 as Phase I program. First, more than 90% of Cs could be recovered from the actual spent fuel [IXC(I) step]. The next is the adsorption of the platinum group metals (PGM), lanthanides, Am and Cm were separated by using a tertiary pyridine-type resin (TPR) as ion exchange steps [IXC(II, III, IV) steps]. The separated PGM metals will be supplied to the electrochemical extraction [CEE step]. As experiment for safety issues, Hastelloy-B at RT and Ta at 90C were confirmed their anti-corrosive in highly concentrated HCl media. Thermo-chemical stability for TPR was verified. Issues to be solved for next phase based on the final results of phase I program.

Optimizing composition of TODGA/SiO-P adsorbent for extraction chromatography process

Watanabe, So; Arai, Tsuyoshi*; Ogawa, Tsuyoshi*; Takizawa, Makoto*; Sano, Kyohei*; Nomura, Kazunori; Koma, Yoshikazu

Procedia Chemistry, 7, p.411 - 417, 2012/00

https://doi.org/10.1016/j.proche.2012.10.064

Chemical durability of iron-phosphate glass as the high level waste from pyrochemical reprocessing

Kofuji, Hirohide; Yano, Tetsuji*; Myochin, Munetaka; Matsuyama, Kanae*; Okita, Takeshi*; Miyamoto, Shinya*

Procedia Chemistry, 7, p.764 - 771, 2012/00

https://doi.org/10.1016/j.proche.2012.10.116

As a part of the research and development for the nuclear waste disposal concept suitable to the advanced fuel cycle systems and its performance evaluation, the iron-phosphate glass is examined as an alternative waste form for high level waste generated from pyrochemical reprocessing. In order to enhance the waste element content in the glass matrix and improve the durability of the waste form, optimization experiments of glass composition were carried out and the effect of additional other transition metal oxides was found out in this study.