Nakahara, Masaumi; Shibata, Atsuhiro
Journal of Nuclear Science and Technology, 10 Pages, 2023/00
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.
Hashikura, Yasuaki*; Ishijima, Yasuhiro; Nakahara, Masaumi; Sano, Yuichi; Ueno, Fumiyoshi; Abe, Hitoshi
Hozengaku, 19(3), p.95 - 102, 2020/10
A plutonium concentrator was selected, and constant load tensile tests with controlled applied potentials and electrochemical tests were conducted in nitric acid and sodium nitrate solutions. From the results, a map which shows the effect of nitric acid concentration to crack initiation potential was drawn. And, it was pointed out that not only the nitric acid but also the nitrate ion coordinated to the nitrate must be considered in evaluating the possibility of stress corrosion cracking.
Nakahara, Masaumi; Sano, Yuichi; Nomura, Kazunori
Radiochimica Acta, 108(9), p.701 - 706, 2020/09
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.
Nakahara, Masaumi; Watanabe, So; Ogi, Hiromichi*; Arai, Yoichi; Aihara, Haruka; Motoyama, Risa; Shibata, Atsuhiro; Nomura, Kazunori; Kajinami, Akihiko*
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.66 - 70, 2019/09
A wide variety of hazardous and radioactive liquid waste has generated derived from an advanced aqueous separation experiments in the Chemical Processing Facility. Therefore, they should be stabilized for the safety handling and management. In this study, we report a precipitation or an oxidation for hazardous materials, a solvent extraction for recovery of nuclear materials, and a concentration of solution by a freeze-drying method.
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.
Nomura, Kazunori; Ogi, Hiromichi*; Nakahara, Masaumi; Watanabe, So; Shibata, Atsuhiro
International Journal of Nuclear and Quantum Engineering (Internet), 13(5), p.209 - 212, 2019/00
Nakahara, Masaumi; Sano, Yuichi; Abe, Hitoshi
Progress in Nuclear Science and Technology (Internet), 5, p.52 - 55, 2018/11
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.
Nakahara, Masaumi; Sano, Yuichi; Nomura, Kazunori; Takeuchi, Masayuki
Journal of Chemical Engineering of Japan, 51(3), p.237 - 242, 2018/03
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.
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.
Takeuchi, Masayuki; Sano, Yuichi; Watanabe, So; Nakahara, Masaumi; Aihara, Haruka; Kofuji, Hirohide; Koizumi, Tsutomu; Mizuno, Tomoyasu
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Takeuchi, Masayuki; Aihara, Haruka; Nakahara, Masaumi; Tanaka, Kotaro*
Procedia Chemistry, 21, p.182 - 189, 2016/12
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.
Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo
Journal of Nuclear Science and Technology, 50(11), p.1062 - 1070, 2013/11
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.
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.
Nakahara, Masaumi; Kaji, Naoya; Yano, Kimihiko; Shibata, Atsuhiro; Takeuchi, Masayuki; Okano, Masanori; Kuno, Takehiko
Journal of Chemical Engineering of Japan, 46(1), p.56 - 62, 2013/01
The influence of HNO concentration in the solution on the formation of CsPu(NO) was evaluated in the U crystallization process. The solubility of CsPu(NO) in a uranyl nitrate solution was found to decrease with increasing HNO concentration in the solution. In the U crystallization experiments with the dissolver solution of irradiated fast reactor fuel, CsPu(NO) formed with 6.5 mol/dm HNO concentration in the mother liquor, and the decontamination factor of Cs for the uranyl nitrate hexahydrate crystals was low. Meanwhile, CsPu(NO) did not precipitate with uranyl nitrate hexahydrate crystals under the condition of 4.0 mol/dm HNO concentration in the mother liquor, and Cs could be separated from the uranyl nitrate hexahydrate crystals.
Nakahara, Masaumi; Nakajima, Yasuo; Koizumi, Tsutomu
Industrial & Engineering Chemistry Research, 51(46), p.15170 - 15175, 2012/11
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.
Nakahara, Masaumi; Nomura, Kazunori
Radiochimica Acta, 100(11), p.821 - 826, 2012/11
A relationship between crystal size and decontamination factor of impurities for uranyl nitrate hexahydrate crystals was examined with a dissolver solution of irradiated fast reactor fuel. The large crystal size reduced the specific surface area of the crystals which in turn decreased the adhesion of liquid impurities on the surface of the crystals. Therefore, high decontamination factors of liquid impurities were achieved. However, the uranyl nitrate hexahydrate crystal size did not affect the decontamination of solid impurities in the experiments.
Nakahara, Masaumi; Nakajima, Yasuo; Koizumi, Tsutomu
Industrial & Engineering Chemistry Research, 51(40), p.13245 - 13250, 2012/10
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.
Nakahara, Masaumi; Kaji, Naoya; Nomura, Kazunori
JAEA-Research 2012-009, 15 Pages, 2012/06
In terms of preventing the formation of Pu and Cs compound, Cs in the feed solution should decrease in the U crystallization process. In order to separate Cs contained within irradiated nuclear fuel, the immersion experiments were carried out with the pure water and diluted HNO solution. The elusion ratio of Cs within the powdered fuel in the pure water and 0.1 mol/dm HNO solution after 67 h was 33.8 and 38.3%, respectively. The experimental results suggest a possible beneficial effect of Cs elusion by immersion of the powdered fuel in the pure water and diluted HNO solution before the fuel dissolution process.
Nakahara, Masaumi; Yano, Kimihiko; Shibata, Atsuhiro; Takeuchi, Masayuki; Okano, Masanori; Kuno, Takehiko
Procedia Chemistry, 7, p.282 - 287, 2012/00
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.
Shibata, Atsuhiro; Yano, Kimihiko; Sambommatsu, Yuji; Nakahara, Masaumi; Takeuchi, Masayuki; Washiya, Tadahiro; Nagata, Masanobu*; Chikazawa, Takahiro*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
JAEA has been developing a U crystallization process. The development targets were DFs of over 100, confirmation of mechanical performance of crystallizer, and so on. Fundamental data were obtained by beaker-scale experiments with actual dissolver solution. DFs for most of the FPs are improved by washing. However the formation of Pu-Cs double salt causes low DF of Cs. To confirm the mechanical performance of an annular type crystallizer and a crystal separator, some experiments were carried out. The crystallizer and the separator have good performance. However washing of UNH crystals by the separator did not have the intended effect for solid impurities. We discussed the application of crystal purification technology to improve the purity and selected KCP. UNH crystal purification tests were carried out using bench-scale KCP apparatus with simulated solid impurities. The purifier has good performance on the decontamination of not only liquid impurities but also solid impurities.