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Journal Articles

Ion beam induced luminescence analysis of europium complexes in styrene-divinylbenzene copolymer-coated spherical silica by proton and argon ion beams irradiation

Nakahara, Masaumi; Watanabe, So; Takeuchi, Masayuki; Yuyama, Takahiro*; Ishizaka, Tomohisa*; Ishii, Yasuyuki*; Yamagata, Ryohei*; Yamada, Naoto*; Koka, Masashi*; Kada, Wataru*; et al.

Nuclear Instruments and Methods in Physics Research B, 542, p.144 - 150, 2023/09

The structures of Eu complexes in the adsorbents prepared with various extractants were evaluated by ion beam induced luminescence analysis in an extraction chromatography for minor actinides recovery. The luminescence of Eu was measured with a proton beam obtained from the single-ended accelerator and an argon ion beam obtained from the azimuthally varying field cyclotron in Takasaki Ion Accelerators for Advanced Radiation Application in National Institutes for Quantum Science and Technology. In this study, it was confirmed that the spectral shape of Eu complexes in the adsorbents varied depending on the kinds of extractants, and the correlation between the change in the spectra and the structures of Eu complexes was investigated.

Journal Articles

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, 60(7), p.849 - 858, 2023/07

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

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.

Journal Articles

Ion beam induced luminescence spectra of europium complexes in silica-based adsorbent

Nakahara, Masaumi; Watanabe, So; Ishii, Yasuyuki*; Yamagata, Ryohei*; Yamada, Naoto*; Koka, Masashi*; Yuyama, Takahiro*; Ishizaka, Tomohisa*; Kada, Wataru*; Hagura, Naoto*

QST-M-39; QST Takasaki Annual Report 2021, P. 62, 2023/02

In an extraction chromatography method for minor actinides recovery, analysis for the structure of complex in a silica-based adsorbent has been studied to separate minor actinides efficiently. In this study, Eu was used as simulated material of minor actinides, and silica-based adsorbents containing Eu complex were prepared. The ion beam induced luminescence spectra of Eu complexes in silica-based adsorbents were measured, and the basic data for the structures of Eu complexes were obtained.

Journal Articles

Harmless treatment of radioactive liquid wastes for safe storage in systematic treatment of radioactive liquid waste for decommissioning project

Nakahara, Masaumi; Watanabe, So; Aihara, Haruka; Takahatake, Yoko; Arai, Yoichi; Ogi, Hiromichi*; Nakamura, Masahiro; Shibata, Atsuhiro; Nomura, Kazunori

Proceedings of International Conference on Nuclear Fuel Cycle; Sustainable Energy Beyond the Pandemic (GLOBAL 2022) (Internet), 4 Pages, 2022/07

Various radioactive wastes have been generated from Chemical Processing Facility for basic research on advanced reactor fuel reprocessing, radioactive waste disposal, and nuclear fuel cycle technology. Many types of reagents have been used for the experiments, and some troublesome materials were produced in the course of experiments. The radioactive liquid wastes were treated for stable and safe storage using decomposition, solvent extraction, precipitation, and solidification methods. In this study, current status of harmless treatment for the radioactive liquid wastes would be reported.

Journal Articles

Consideration on stress corrosion cracking evaluation of zirconium for Fuel Reprocessing Facilities

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.

Journal Articles

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

 Times Cited Count:0 Percentile:0.01(Chemistry, Inorganic & Nuclear)

To evaluate the corrosion behavior of a Pu evaporator made from Zr in a reprocessing plant, the influence of PuO$$_{2}$$$$^{2+}$$ 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$$^{-3}$$ HNO$$_{3}$$. However, there were no significant changes at high PuO$$_{2}$$$$^{2+}$$ concentrations, and Zr showed high corrosion resistance under our experimental conditions.

Journal Articles

Stabilization processing of hazardous and radioactive liquid wastes derived from advanced aqueous separation experiments for safety handling and management of waste

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.

Journal Articles

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.

Journal Articles

Waste management in a Hot Laboratory of Japan Atomic Energy Agency, 1; Overview and activities in chemical processing facility

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

Journal Articles

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

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$$_{3}$$ concentrations and temperature. However, these experimental results imply that Zr has high corrosion resistance in Pu nitrate solutions.

Journal Articles

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

 Times Cited Count:2 Percentile:9.15(Engineering, Chemical)

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.

Journal Articles

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.

Journal Articles

Minor actinides recovery from irradiated fuel for SmART cycle test

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

Journal Articles

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

 Times Cited Count:1 Percentile:81.34

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.

Journal Articles

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

 Times Cited Count:4 Percentile:32.63(Nuclear Science & Technology)

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$$^{3}$$ HNO$$_{3}$$ was supplied at 21$$^{circ}$$C 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.

Journal Articles

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$$_{3}$$ 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.

Journal Articles

Nitric acid concentration dependence of dicesium plutonium(IV) nitrate formation during solution growth of uranyl nitrate hexahydrate

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

 Times Cited Count:1 Percentile:6.34(Engineering, Chemical)

The influence of HNO$$_{3}$$ concentration in the solution on the formation of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ was evaluated in the U crystallization process. The solubility of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in a uranyl nitrate solution was found to decrease with increasing HNO$$_{3}$$ concentration in the solution. In the U crystallization experiments with the dissolver solution of irradiated fast reactor fuel, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ formed with 6.5 mol/dm$$^{3}$$ HNO$$_{3}$$ concentration in the mother liquor, and the decontamination factor of Cs for the uranyl nitrate hexahydrate crystals was low. Meanwhile, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ did not precipitate with uranyl nitrate hexahydrate crystals under the condition of 4.0 mol/dm$$^{3}$$ HNO$$_{3}$$ concentration in the mother liquor, and Cs could be separated from the uranyl nitrate hexahydrate crystals.

Journal Articles

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

 Times Cited Count:2 Percentile:11.13(Engineering, Chemical)

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$$_{3}$$ 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$$_{3}$$ solution.

Journal Articles

Effect of crystal size on purity of uranyl nitrate hexahydrate crystalline particles grown in nitric acid medium

Nakahara, Masaumi; Nomura, Kazunori

Radiochimica Acta, 100(11), p.821 - 826, 2012/11

 Times Cited Count:0 Percentile:10.14(Chemistry, Inorganic & Nuclear)

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.

Journal Articles

Extraction behavior of fission products with tri-${it n}$-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

 Times Cited Count:7 Percentile:27.09(Engineering, Chemical)

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-${it n}$-butyl phosphate. The experimental results show that Zr was removed with a low HNO$$_{3}$$ concentration scrubbing solution and Tc was decontaminated using a high HNO$$_{3}$$ concentration solution. Other fission products were effectively decontaminated in the system.

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