Outline of Japan Atomic Energy Agency's Okuma Analysis and Research Center, 3; Laboratory-2

Ito, Masayasu; Ogawa, Miho; Inoue, Toshihiko; Yoshimochi, Hiroshi; Koyama, Shinichi; Koyama, Tomozo; Nakayama, Shinichi

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 7 Pages, 2017/00

http://hotlab.sckcen.be/proceedings

Laboratory-2 of the Okuma Analysis and Research Center will be used for the technological development of techniques to treat and dispose fuel debris, etc. The specific analytical content and its importance has been discussed by an experts committee in FY 2016. The committee regarded fuel debris retrieval and criticality control related topics as the most important content. As a result, it will be a priority to introduce equipment to perform examination such as shape and size measurement, compositional and nuclide analysis, hardness and toughness test, and radiation dose rate measurement. In addition, since sample will have high dose rates (1 Sv/h or more) at the time of reception, hot cells with enough radiation shielding ability will be used. In the hot cell, the pre-processing will be performed, such as cutting and dissolution of samples. Processed samples will be examined in concrete cells, steel cells, glove boxes and fume hoods. Detail design of Laboratory-2 started on FY 2017.

Program of the analysis and research laboratory for Fukushima-Daiichi and advanced techniques to be applied in the laboratory

Sekio, Yoshihiro; Yoshimochi, Hiroshi; Kosaka, Ichiro; Hirano, Hiroyasu; Koyama, Tomozo; Kawamura, Hiroshi

Proceedings of 52nd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2015) (Internet), 8 Pages, 2015/09

http://hotlab.sckcen.be/en/Proceedings

Due to the Fukushima Daiichi Nuclear Power Plant accident in March 2011, the safe and secure implementations of the decommissioning for Fukushima Daiichi Nuclear Power Plant has been positioned as the urgent tasks in Japan. Japan Atomic Energy Agency has a critical mission of analysing radioactive wastes having generated by the accident for long-term storage and disposal methods. This will be performed in two hot laboratories to be constructed in Okuma Analysis and Research Center at Fukushima Daiichi Nuclear Power Plant site. In one laboratory, radioactive wastes such as rubbles and secondary wastes will be treated, whereas debris such as fuel debris and high dose structural materials will be handled in the other laboratory. The detail considerations for advanced techniques and experimental apparatus to be installed are underway.

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

Batch crystallization of uranyl nitrate

Chikazawa, Takahiro*; Kikuchi, Toshiaki*; Shibata, Atsuhiro; Koyama, Tomozo; Homma, Shunji*

Journal of Nuclear Science and Technology, 45(6), p.582 - 587, 2008/06

https://doi.org/10.3327/jnst.45.582

Batch crystallization of uranyl nitrate is carried out in order to obtain fundamental data required for the development of reprocessing involving crystallization. Particular attention is paid to the development of a method for predicting the concentrations of uranium and nitric acid in the mother liquor and the amount of uranyl nitrate crystals produced. Initial concentrations of uranyl nitrate and nitric acid are 500-600 g/dm and 4-6 mol/dm, respectively, corresponding to the condition of a dissolver solution of spent fuel. Steady-state mass balance equations including the correlation equation for the equilibrium solubility of uranium nitrate are applied to the prediction. The calculated concentrations of uranium and nitric acid are in close agreement with the experimental ones. The recovery of uranium is accurately predicted by the calculated concentrations, with an error of less than 5%.

Development of actinides co-extraction system with direct extraction process using super-critical fluid

Koyama, Tomozo

Genshiryoku eye, 53(9), p.56 - 59, 2007/09

http://ci.nii.ac.jp/naid/40015601668

In order to establish improvement of aqueous reprocessing technology in economy and waste reduction, JAEA has been developing "Actinides co-extraction system with direct extraction process using supercritical carbon dioxide" in collaboration with Nagoya University and ARTECH Co.Ltd. as nuclear system research and development application program by MEXT since 2005. Measurement of distribution factors of major fission product elements and confirmation of supercritical fluid behavior have been executed. In JFY 2007, distribution factors will be continuously measured and direct extraction test of unirradiated MOX in normal pressure will be carried out. In JFY 2008-2009, direct extraction test of unirradiated MOX and irradiated fuel will be executed using supercritical carbon dioxide in a glovebox and a cell.

Separation of actinide elements by solvent extraction using centrifugal contactors in the NEXT process

Nakahara, Masaumi; Sano, Yuichi; Koma, Yoshikazu; Kamiya, Masayoshi; Shibata, Atsuhiro; Koizumi, Tsutomu; Koyama, Tomozo

Journal of Nuclear Science and Technology, 44(3), p.373 - 381, 2007/03

https://doi.org/10.3327/jnst.44.373

Actinides recovery was attempted by the simplified solvent extraction process using TBP as an extractant for U, Pu and Np co-recovery and the SETFICS process for Am and Cm recovery with a view to decreasing the environmental impact. Uranium, Pu and Np co-recovery was conducted under the condition with high nitric acid concentration in the feed solution or scrubbing solution. High nitric acid concentration in the feed solution availed to the Np oxidation not only in the feed solution, but also at the extraction section. This oxidation reaction permitted the Np extraction with U and Pu. In the SETFICS process, a TRUEX solvent of 0.2M CMPO/1.4M TBP was employed to increase the loading of metals. In place of sodium nitrate, HAN was applied to this experimental flow sheet for "salt-free" concept. This experiment was succeeded in Am and Cm product. On high-loading flow sheet, the flow rate of aqueous effluents and spent solvent was expected to decrease in 47% and 54%, respectively.

Crystallization behavior of uranium and plutonium in nitric acid solution

Yano, Kimihiko; Shibata, Atsuhiro; Nomura, Kazunori; Koizumi, Tsutomu; Koyama, Tomozo

Recent Advances in Actinide Science, p.644 - 646, 2006/06

U crystallization is effective to minimize solvent extraction process equipment by separating U from dissolver solution of spent fuel. In order to establish crystallization process for FBR MOX fuel, Pu behavior needs to be investigated with mixed U and Pu solution. Pu valence influence on the decontamination factor (DF) of Pu to U in crystal. Pu(IV) was not crystallized with U and DF was about 30. Although Pu(VI) dose not solely crystallized, it was co-crystallized with U and DF of Pu to U in crystal was about 1.3. The effect of Pu(IV) concentration on crystallization yield of U is small. It had a tendency to decrease crystallization yield of U by that Pu(VI) exists, compared with that calculated by solubility of U in U-HNO-HO system.

Uranium crystallization for dissolver solution of irradiated FBR MOX fuel

Yano, Kimihiko; Shibata, Atsuhiro; Nomura, Kazunori; Koizumi, Tsutomu; Koyama, Tomozo

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10

The uranium crystallization process has been developed as part of advanced aqueous reprocessing system. In the process, the great part of uranium is separated from the dissolver solution by crystallization as uranyl nitrate hydrate. The uranium crystallization test was carried out with the dissolver solution of irradiated FBR MOX fuel to investigate the influence of cooling rate on the crystal size and the behavior of fission products compared with that of tetravalent plutonium with the real dissolver solution. In regard to the influence of the cooling rate, it was obvious that the crystal size was smaller as the cooling ratio was faster. However, it was suggested that a large crystal produced by crystallization was always not high purity. Concerning the behavior of fission products, europium behaved similarly to tetravalent plutonium. Cesium accompanied with uranium into the crystals under the condition of this test.

U(VI) back-extraction trials for measurement of U(VI) mass transfer efficiency in single stage centrifugal contactor

Sano, Yuichi; Shibata, Atsuhiro; Koizumi, Tsutomu; Koyama, Tomozo; Fox, D.*; Carrott, M.*; Taylor, R.*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10

Based on the U(VI) back-extraction experiments with the miniaturized single stage centrifugal contactor, effects of the total flowrate, the rotor speed and the temperature on the U(VI) mass transfer efficiency were discussed. The relationship between the U(VI) mass transfer efficiency and the residence time suggested that the U(VI) back-extraction system with the centrifugal contactor could be treated as a perfectly stirred continuous reactor. It is considered that the variations of the mass transfer efficiency were caused by the flow pattern change from the laminar flow to the turbulent flow with increasing of the rotor speed in the centrifugal contactor. In the U(VI) back-extraction under the high rotor speed condition, no significant temperature dependence of the apparent overall transform coefficient was observed. In the U(VI) back-extraction under the low rotor speed condition, temperature dependence of the apparent overall transform coefficient seemed to be larger than under the high rotor speed condition, and it suggested a chemical-step contribution in this system. Such a difference of the temperature dependence of the apparent overall transform coefficient between these experiments might be caused by the change of solvent extraction environments in the centrifugal contactor with the variation of the rotor speed.

Direct Extraction of Uranium and Plutonium from Oxide Fuel using TBP-HNOComplex for Super-DIREX Process

Kamiya, Masayoshi; Miura, Sachiko; Nomura, Kazunori; Koyama, Tomozo; Ogumo, Shinya*; Mori, Yukihide*; Enokida, Yoichi*

CD-ROM, P1-35, 4P., 4 Pages, 2004/00

Super-DIREX is a new reprocessing method which has high economical efficiency. Experimental study of this process was started on the direct extraction of U and Pu from irradiated MOX fuel by the supercritical carbon dioxide (SFCO) containing TBP-HNO complex. This report describes direct extraction of U and Pu with TBP-HNO3 complex at atmospheric pressure, as the first test for irradiated fuel, in order to investigate the applicability of SFCO containing TBP-HNO complex. In this test, dependency on dissolution temperature, Pu content, fuel/ TBP-HNO complex ratio and effect of voloxidation were investigated. From these results, TBP-HNO complex was found to be effective in the respect of the recovery of U and Pu. The number of the process step in dissolution and co-extraction is small, and amount of waste can be reduced. It is applicable to the direct extraction in Super-DIREX.