Influence of plutonium content in dissolver solutions derived from irradiated fast reactor fuels on plutonium stripping in multistage countercurrent liquid-liquid extraction with acid split flowsheet

Nakahara, Masaumi; Shibata, Atsuhiro

Journal of Nuclear Science and Technology, 10 Pages, 2023/00

https://doi.org/10.1080/00223131.2022.2162137

To develop the acid split method which has highly nuclear proliferation resistance, influence of Pu content in dissolver solutions derived from irradiated fast reactor fuel on the Pu stripping was investigated in experiments and a calculation. The Pu content in the U/Pu and U products increased with increasing the Pu content in the dissolver solution. Moreover, the calculated results indicate that the Pu leakage into the U product is suppressed with the Pu stripping solution only at low temperature.

Influence of plutonyl ion on electrochemical characterization of zirconium in plutonium nitrate solutions

Nakahara, Masaumi; Sano, Yuichi; Nomura, Kazunori

Radiochimica Acta, 108(9), p.701 - 706, 2020/09

https://doi.org/10.1515/ract-2019-3228

To evaluate the corrosion behavior of a Pu evaporator made from Zr in a reprocessing plant, the influence of PuO was investigated with Pu nitrate solutions in electrochemical experiments. The maximum open circuit potential of Zr in the Pu nitrate solution was approximately 1 V in the Pu nitrate solution containing 7 mol dm HNO. However, there were no significant changes at high PuO concentrations, and Zr showed high corrosion resistance under our experimental conditions.

Waste management in a Hot Laboratory of Japan Atomic Energy Agency, 3; Volume reduction and stabilization of solid waste

Nakahara, Masaumi; Watanabe, So; Ogi, Hiromichi*; Shibata, Atsuhiro; Nomura, Kazunori

International Journal of Nuclear and Quantum Engineering (Internet), 13(4), p.191 - 194, 2019/04

High level radioactive solid waste is reduced the volume or stabilized in the Chemical Processing Facility in the Japan Atomic Energy Agency. A plastic product is molten with a heating mantle and reduced the volume. A non-flammable such as metal is cut with a band saw machine for reducing the volume. A used adsorbent in the extraction chromatograph process was heated with an electric furnace using non-radioactive materials, and the experimental result suggests that organic materials in the used adsorbent were decomposed stably.

Electrochemical properties of zirconium in highly concentrated plutonium nitrate solution

Nakahara, Masaumi; Sano, Yuichi; Abe, Hitoshi

Progress in Nuclear Science and Technology (Internet), 5, p.52 - 55, 2018/11

https://doi.org/10.15669/pnst.5.52

For evaluating the secular change of Pu evaporator made of Zr in the commercialized nuclear fuel reprocessing plant, electrochemical experiments were carried out with Pu nitrate solutions. The open circuit potentials of Zr increased with increasing Pu, HNO concentrations and temperature. However, these experimental results imply that Zr has high corrosion resistance in Pu nitrate solutions.

Actinides recovery from irradiated fuel for SmART cycle

Sano, Yuichi; Watanabe, So; Nakahara, Masaumi; Aihara, Haruka; Takeuchi, Masayuki

Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 4 Pages, 2017/09

JAEA has been promoting MA recycle project using a FR fuel cycle named as SmART cycle concept. The SmART cycle contains the recovery of all actinides, in which total amount of MA is estimated to around 1-2g, at CPF from the FR Joyo spent fuel, the fabrication of MA bearing MOX fuel pellets and pins at AGF with recovered actinides, and the irradiation test of the fabricated fuels at the Joyo. In this paper, recent activities on actinides recovery in CPF, which will make a significant contribution to the SmART cycle, were summarized.

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.

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.

Extraction behavior of fission products with tri--butyl phosphate by countercurrent multistage extraction in a uranium, plutonium, and neptunium co-recovery system

Nakahara, Masaumi; Nakajima, Yasuo; Koizumi, Tsutomu

Industrial & Engineering Chemistry Research, 51(40), p.13245 - 13250, 2012/10

https://doi.org/10.1021/ie3011287

To study the extraction behavior of fission products, the countercurrent multistage experiments were carried out with centrifugal contactors in the U, Pu, and Np co-recovery system. Neptunium was co-recovered with U and Pu using tri--butyl phosphate. The experimental results show that Zr was removed with a low HNO concentration scrubbing solution and Tc was decontaminated using a high HNO concentration solution. Other fission products were effectively decontaminated in the system.

Characteristics of dicesium plutonium(IV) nitrate formation in separation system of uranyl nitrate hexahydrate crystal

Nakahara, Masaumi; Yano, Kimihiko; Shibata, Atsuhiro; Takeuchi, Masayuki; Okano, Masanori; Kuno, Takehiko

Procedia Chemistry, 7, p.282 - 287, 2012/00

https://doi.org/10.1016/j.proche.2012.10.045

For decontamination of Cs and Pu compound, CsPu(NO), precipitated in the U cooling crystallization method, solubility measurement of CsPu(NO) in a uranyl nitrate solution and a U crystallization experiments were carried out with the dissolver solution derived from irradiated fast neutron reactor core fuel. The solubility of CsPu(NO) in the uranyl nitrate solution decreased with decreasing temperature. In the crystallization experiments, the decontamination factors of Cs and Pu for uranyl nitrate hexahydrate crystal decrease with increasing the Cs concentration in the feed solution because CsPu(NO) formed in the course of U crystallization. Basic data were obtained for the formation behavior of CsPu(NO) in the U crystallization process.