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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:1 Percentile:8.18(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

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:71.18

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:3 Percentile:29.69(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:7.24(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:12.43(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:0.01(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:4 Percentile:20.39(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.

JAEA Reports

Elution properties of cesium contained within irradiated fast neutron reactor fuel in water and diluted nitric acid solution

Nakahara, Masaumi; Kaji, Naoya; Nomura, Kazunori

JAEA-Research 2012-009, 15 Pages, 2012/06

JAEA-Research-2012-009.pdf:6.37MB

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$$_{3}$$ solution. The elusion ratio of Cs within the powdered fuel in the pure water and 0.1 mol/dm$$^{3}$$ HNO$$_{3}$$ 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$$_{3}$$ solution before the fuel dissolution process.

Journal Articles

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

 Times Cited Count:1 Percentile:60.25

For decontamination of Cs and Pu compound, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$, precipitated in the U cooling crystallization method, solubility measurement of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ 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 Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ 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 Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ formed in the course of U crystallization. Basic data were obtained for the formation behavior of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in the U crystallization process.

Journal Articles

FaCT Phase I evaluation on the advanced aqueous reprocessing process, 5; Research and development of uranium crystallization system

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.

Journal Articles

Behavior of fission products in simplified solvent extraction system for uranium, plutonium and neptunium co-recovery

Nakahara, Masaumi; Shibata, Atsuhiro; Koma, Yoshikazu

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 4 Pages, 2011/12

To elucidate the behavior of fission products, two counter current experiments were carried out with a dissolver solution derived from irradiated fast neutron reactor "JOYO" core fuel, one was a single cycle flow sheet using double scrubbing solution with 9 and 1 mol/dm$$^{3}$$ HNO$$_{3}$$ and the other used a Tc scrubbing solution with 10 mol/dm$$^{3}$$ HNO$$_{3}$$. Among fission products, the decontamination behavior of Zr and Tc differed according to HNO$$_{3}$$ concentration of the scrubbing solutions. The decontamination factor of Zr and Tc increase to $$>$$76.8 and $$>$$7.52 with 10 mol/dm$$^{3}$$ HNO$$_{3}$$ in the Tc scrubbing solution. Other fission products such as Cs was well decontaminated and its DF resulted in 10$$^{5}$$.

Journal Articles

FaCT Phase-I evaluation on the advanced aqueous reprocessing process, 4; Solvent extraction simplified for FBR fuel reprocessing

Koma, Yoshikazu; Ogino, Hideki; Sakamoto, Atsushi; Nakabayashi, Hiroki; Shibata, Atsuhiro; Nakahara, Masaumi; Washiya, Tadahiro

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12

Journal Articles

Influence of nitric acid and nitrous acid on oxidation and extraction of neptunium with double scrub flow sheet in simplified solvent extraction process

Nakahara, Masaumi; Koma, Yoshikazu

Journal of Chemical Engineering of Japan, 44(5), p.313 - 320, 2011/05

 Times Cited Count:2 Percentile:12.02(Engineering, Chemical)

The influence of the HNO$$_{3}$$ concentration in the feed and scrubbing solutions on the behavior of Np was evaluated experimentally and found to be co-extracted into the tri-${it n}$-butylphosphate with U and Pu. Almost all the Np in a 4.9 mol/dm$$^{3}$$ HNO$$_{3}$$ feed solution was recovered with U and Pu, based on the experimental flow sheet with double scrubbing solutions of 9 and 1 mol/dm$$^{3}$$ HNO$$_{3}$$. The experimental results showed a large contribution from the HNO$$_{3}$$ concentration in the feed solution and at the extraction section to Np(V) oxidation. On the other hand, calculation results showed that high HNO$$_{2}$$ concentrations in the feed solution tended to leak Np into the raffinate.

59 (Records 1-20 displayed on this page)