Quantitative analysis of Zr adsorbed on IDA chelating resin using Micro-PIXE

Arai, Yoichi; Watanabe, So; Ono, Shimpei; Nomura, Kazunori; Nakamura, Fumiya*; Arai, Tsuyoshi*; Seko, Noriaki*; Hoshina, Hiroyuki*; Kubota, Toshio*

QST-M-23; QST Takasaki Annual Report 2018, P. 59, 2020/03

http://id.nii.ac.jp/1657/00079235/

A Review of separation processes proposed for advanced fuel cycles based on technology readiness level assessments

Baron, P.*; Cornet, S. M.*; Collins, E. D.*; DeAngelis, G.*; Del Cul, G.*; Fedorov, Y.*; Glatz, J. P.*; Ignatiev, V.*; Inoue, Tadashi*; Khaperskaya, A.*; et al.

Progress in Nuclear Energy, 117, p.103091_1 - 103091_24, 2019/11

https://doi.org/10.1016/j.pnucene.2019.103091

The results of an international review of separation processes for spent nuclear fuel (SNF) recycling in future closed fuel cycles with the evaluation of Technology Readiness Level are reported. This study was made by the Expert Group on Fuel Recycling Chemistry (EGFRC) organised by the Nuclear Energy Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD). A unique feature of this study was that processes were classified according to a hierarchy of separations aimed at different elements within spent fuel (uranium; uranium-plutonium co-recovery; minor actinides; high heat generating radionuclides) and also the Head-end processes, used to prepare the SNF for chemical separation, were included. Separation processes covered both wet (hydrometallurgical) and dry (pyro-chemical) processes.

Partitioning of plutonium by acid split method with dissolver solution derived from irradiated fast reactor fuel with high concentration of plutonium

Nakahara, Masaumi; Sano, Yuichi; Nomura, Kazunori; Takeuchi, Masayuki

Journal of Chemical Engineering of Japan, 51(3), p.237 - 242, 2018/03

https://doi.org/10.1252/jcej.17we264

For evaluating the Pu partitioning behavior under the condition of high Pu concentration in the feed solution by the acid split method, the counter current experiment was carried out. The Pu content in the U/Pu product was 1.51 times higher than that in the feed solution. In the Pu partitioning section, Pu polymerization and third phase formation were observed, and the operation of centrifugal contactors was stable.

Research of process to treat the radioactive liquid waste containing chloride ion generated by pyroprocessing plant in operating

Tada, Kohei; Kitawaki, Shinichi; Watanabe, So; Aihara, Haruka; Shibata, Atsuhiro; Nomura, Kazunori

Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 3 Pages, 2017/09

Radioactive liquid waste containing chloride ion (Cl) is generated by chemical analysis for process control of pyroprocessing. To realize discharging this liquid waste to the sea, it's necessary to carry out the process in order to separate Cl and recover U, Pu. This study carried out a combination of the AgCl precipitation method and extraction chromatography method to separate Cl and recover U, Pu. The result of precipitation test showed that U and Pu didn't occur the co-precipitation after the test. The result of solid phase extraction test showed that 95% of Pu was successfully recovered from the liquid waste. It was difficult to analyze radioactivity about U because the concentration of U is not enough. These results showed that these process has the feasibility of the discharging the liquid waste to the sea.

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

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.

Chemical composition of insoluble residue generated at the Rokkasho Reprocessing Plant

Yamagishi, Isao; Odakura, Makoto; Ichige, Yoshiaki; Kuroha, Mitsuhiko; Takano, Masahide; Akabori, Mitsuo; Yoshioka, Masahiro*

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1113 - 1119, 2015/09

The characteristics of insoluble residues in fine suspension at the Rokkasho Reprocessing Plant were analyzed. The insoluble residues were washed with oxalic acid solution to dissolve zirconium molybdate residues. XRD profiles of unwashed residues showed the presence of a noble metal alloy, zirconium molybdate, and zirconia, but zirconium molybdate was not found after washing. More than 50% of the Sb-125 and Pu in thee residues was washed out as well. The noble metal alloy composed of Mo, Tc, Ru, Rh, and Pd occupied more than 90% of the total weight of 12 elements (Ca, Cr, Fe, Ni, Zr, Mo, Tc, Ru, Rh, Pd, Te, and U) found in the residues. In consideration of the chemical forms of 12 elements, the alloy-to-residue weight ratio was evaluated to be 64% and 78% with and without 18% of an unknown component, respectively.

Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo

Journal of Nuclear Science and Technology, 50(11), p.1062 - 1070, 2013/11

https://doi.org/10.1080/00223131.2013.831330

The acid split method for Pu partitioning without reductant was investigated for improving nuclear proliferation resistance, safety, and cost. A practical acid split flow sheet was configured using a extraction calculation code, and countercurrent experiment was carried out based on their calculation results. 0.15 mol/dm HNO was supplied at 21C for the Pu stripping. The Pu content of the U/Pu product increased to 2.28 times larger than that of the feed solution. In addition, the Pu leakage to the U product was 0.47%. The experimental results indicate that the proposed flow sheet is effective for fast reactor fuel reprocessing.

Plutonium partitioning in uranium and plutonium co-recovery system for fast reactor fuel recycling with enhanced nuclear proliferation resistance

Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.539 - 542, 2013/09

In order to develop a fast reactor fuel reprocessing, countercurrent extraction experiments for Pu reduction partitioning method with hydroxylamine nitrate and acid split method without Pu reductant were carried out. In the Pu reduction method, a part of U was co-recovered with Pu because the U scrubbing part can be deleted in the Pu partitioning section. On the other hand, acid split method experimental results indicate that almost all Pu was recovered with U by supplying diluted HNO solution in the Pu partitioning section. This study shows that not only Pu reduction partitioning but also acid split methods are effective for fast reactor fuel reprocessing.

Washing of uranyl nitrate hexahydrate crystals with nitric acid aqueous solution to improve crystal quality

Nakahara, Masaumi; Nakajima, Yasuo; Koizumi, Tsutomu

Industrial & Engineering Chemistry Research, 51(46), p.15170 - 15175, 2012/11

https://doi.org/10.1021/ie3013067

In the crystal washing experiment using the uranyl nitrate solution containing Ce, the Ce in the mother liquor was attached to the surface of the uranyl nitrate hexahydrate crystals and tend to be removed by the washing operation with low HNO concentration washing solution. In the crystallization experiments using the dissolver solution of irradiated fast reactor core fuel, the decontamination factors of liquid impurities were improved by crystal washing. On the other hand, the decontamination factors of solid impurities decreased with several washings because uranyl nitrate hexahydrate is more soluble than the solid impurities in an HNO solution.