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Hayashi, Hirokazu; Tsubata, Yasuhiro; Sato, Takumi
Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(3), p.97 - 107, 2023/08
The Japan Atomic Energy Agency has chosen nitride fuel as the first candidate for the transmutation of long-lived minor actinides (MA) using accelerator-driven systems (ADS). The pyrochemical method has been considered for reprocessing spent MA nitride fuels, because their decay heat should be very large for aqueous reprocessing. This study was conducted to investigate the effect of decay heat on the pyrochemical reprocessing of MA nitride fuels. On the basis of the estimated decay heats and the temperature limits of the materials that are to be handled in pyrochemical reprocessing, quantities adequate for handling in argon gas atmosphere were evaluated. From these considerations, we proposed that an electrorefiner with a diameter of 26 cm comprising 12 cadmium (Cd) cathodes with a diameter of 4 cm is suitable. On the basis of the size of the electrorefiner, the number necessary to reprocess spent MA fuels from 1 ADS in 200 days was evaluated to be 25. Furthermore, the amount of Cd-actinides (An) alloy to produce An nitrides by the nitridation-distillation combined reaction process was proposed to be about one-quarter that of Cd-An cathode material. The evaluated sizes and required numbers of equipment support the feasibility of pyrochemical reprocessing for MA nitride fuels.
O
) by mechanochemical treatmentAkiyama, Daisuke*; Mishima, Tomoki*; Okamoto, Yoshihiro; Kirishima, Akira*
High Temperature Materials and Processes, 42(1), p.20220268_1 - 20220268_9, 2023/04
A powder mixture of UO and TiO was mechanochemically treated in a planetary ball mill under Ar atmosphere for 1 h using a tungsten carbide vial and balls as the milling medium. Such mechanochemical (MC) treatment reduced the crystallinity of UO and TiO. The mechanochemically treated powder mixture was heated at 973-1573K for 6 h under Ar atmosphere and analyzed by X-ray diffraction analysis, scanning electron microscopy-energy-dispersive X-ray spectroscopy, and X-ray absorption fine structure analysis. UTiO did not form below 1373K without MC treatment and only the starting materials were observed. At 1473 and 1573K, a small amount of UTiO and equal amounts of UTiO and UO were formed, respectively. The mechanochemically treated sample produced nearly pure UTiO containing small amounts of UO impurities when heated above 1173K for 6 h. UTiO was highly crystalline and uniform regardless of the synthesis temperature.
Nuclear Science and Engineering Center; Fuel Cycle Design Office; Plutonium Fuel Development Center; Nuclear Plant Innovation Promotion Office; Fast Reactor Cycle System Research and Development Center; J-PARC Center
JAEA-Review 2022-052, 342 Pages, 2023/02
JAEA-Review-2022-052.pdf:18.05MB
This report summarizes the current status and future plans of research and development (R&D) on partitioning and transmutation technology in Japan Atomic Energy Agency, focusing on the results during the 3rd Medium- to Long-term Plan period (FY 2015-2021). Regarding the partitioning technology, R&D of the solvent extraction method and the extraction chromatography method are described, and regarding the minor actinide containing fuel technology, R&D of the oxide fuel production using the simplified pellet method, the nitride fuel production using the external gelation method, and pyrochemical reprocessing of the nitride fuel were summarized. Regarding transmutation technology, R&D of technology using fast reactors and accelerator drive systems were summarized. Finally, the new facilities necessary for the future R&D were mentioned.
(M = Gd, Np) by the reaction of MN with Pd and chlorination of MPd using cadmium chlorideHayashi, Hirokazu; Shibata, Hiroki; Sato, Takumi; Otobe, Haruyoshi
Journal of Radioanalytical and Nuclear Chemistry, 332(2), p.503 - 510, 2023/02
Times Cited Count：0 Percentile：0.01(Chemistry, Analytical)The formation of MPd (M = Gd, Np) by the reaction of MN with Pd at 1323 K in Ar gas flow was observed. Cubic AuCu
-type GdPd
(
= 0.4081 
0.0001 nm) and NpPd ( = 0.4081 0.0001 nm) were identified, respectively. The product obtained from the reaction of NpN with Pd contained additional phases including the hexagonal TiNi-type NpPd. Chlorination of the MPd (M = Gd, Np) samples was accomplished by the solid-state reaction using cadmium chloride at 673 K in a dynamic vacuum. Pd-rich solid solution phase saturated with Cd and an intermetallic compound PdCd were obtained as by-products of MCl formation.
Murakami, Tsuyoshi*; Hayashi, Hirokazu
Journal of Nuclear Materials, 558, p.153330_1 - 153330_7, 2022/01
Times Cited Count：0 Percentile：0.01(Materials Science, Multidisciplinary)Excess amounts of dissolution agents, CdCl
and ZrCl
, are required to dissolve transuranium (TRU: Pu and minor actinides) nitrides into LiCl-KCl melts at the chemical dissolution step, which is the first step in the reprocessing of used nitride fuels. We propose an electrochemical process where the remaining Zr and Cd are recovered from the melts to be recycled as dissolution agents for the chemical dissolution step, leaving TRU in the melts. Since the initial concentration ratio of CdCl/ZrCl remaining in the melts would depend on the condition of the chemical dissolution step and would vary during the proposed electrochemical recovery process, electrochemical behaviors of Zr and Cd were investigated in LiCl-KCl melts with various concentration ratios of CdCl/ZrCl at 723 K to confirm the basic feasibility of the proposed process. Potentiostatic electrolysis was performed using a liquid Cd cathode at -1.05 V (vs. Ag/AgCl), which was a more positive potential than the redox potentials of TRU on the liquid Cd electrode. The obtained results showed that the current efficiency for recovering Zr and Cd from the melts was as high as 100% regardless of the CdCl/ZrCl concentration ratio in the melts.
Yamamoto, Kazuya; Makino, Takayoshi; Iso, Hidetoshi; Segawa, Tomoomi; Kawaguchi, Koichi; Ishii, Katsunori
JAEA-Technology 2021-002, 31 Pages, 2021/05
JAEA-Technology-2021-002.pdf:4.37MB
In the MOX fuel fabrication process, a dry recycle technology has been developed to effectively utilize dry recovered powder obtained by crushing out of specification MOX pellets. The particle size of the dry recovery powder is divided into three classes; coarse size (about 250 
m or less), medium size (about 100 m or less), and fine size (about 10 m or less) by the current crushers, and the effect of controlling the density of sintered pellets is obtained to a certain extent by adding the dry recovered powder to the raw powder. In this report, with the aim of more finely adjusting the particle size of the dry recovery powder, a buhrstone mill and a collision plate-type jet mill were selected as grinders that can adjust the dry recovered powder within a particle size range of 250 m or less, and the particle size adjustment test was conducted to pulverize the tungsten-carbide-cobalt (WC-Co) pellets as a simulated material for the MOX pellets. The buhrstone mill can control the particle size within a certain range by adjusting the grindstone clearance, but particles with a particle size of 250 m or more may be discharged. On the contrary, it is expected that the particle size of the collision plate-type jet mill can be controlled in the range of 250 m or less by adjusting the classification zone clearance. Therefore, the collision plate-type jet mill is more suitable for adjusting the particle size of the dry recovered powder than the buhrstone mill.
Kudo, Hideyuki*; Otani, Yuichi*; Hara, Masahide*; Kato, Atsushi; Otaka, Masahiko; Ide, Akihiro*
Journal of Nuclear Science and Technology, 57(4), p.408 - 420, 2020/04
Times Cited Count：1 Percentile：10.81(Nuclear Science & Technology)In a fuel handling system of sodium-cooled fast reactors (SFRs), it is necessary to remove the sodium remaining on spent fuel assemblies (FAs) before storing them in a spent fuel water pool (SFP) in order to minimize plant operating loads. A next-generation SFR in Japan has adopted an advanced dry cleaning process which consists of the following steps, argon gas blowing to remove the metallic residual sodium on the FA, moist argon gas blowing to deactivate the residual sodium, and direct storage in the SFP. This three-step process increases economic competitiveness and reduces waste products thanks to a waterless process. In this R&D work, performance of the dry cleaning process has been investigated.
Tateno, Haruka; Sato, Takumi; Tsubata, Yasuhiro; Hayashi, Hirokazu
Journal of Nuclear Science and Technology, 57(3), p.224 - 235, 2020/03
Times Cited Count：6 Percentile：54.98(Nuclear Science & Technology)Fuel cycle technology for the transmutation of long-lived minor actinides (MAs) using an accelerator-driven system has been developed using the double-strata fuel cycle concept. A mononitride solid solution of MAs and Pu diluted with ZrN is a prime fuel candidate for the accelerator-driven transmutation of MAs. Pyro-reprocessing is suitable for recycling the residual MAs in irradiated nitride fuel with high radiation doses and decay heat. Spent nitride fuel is anodically dissolved, and the actinides are recovered simultaneously into a liquid cadmium cathode via molten salt electrorefining. The process should be designed to achieve the target recovery yield of MAs and the acceptable impurity level of rare earths in the recovered material. We evaluated the material balance during the pyro-reprocessing of spent nitride fuel to gain important insight on the design process. We examined the effects of changing processing conditions on material flow and quantity of waste.
Kudo, Hideyuki*; Inuzuka, Taisuke*; Hara, Masahide*; Kato, Atsushi; Nagai, Keiichi; Ide, Akihiro*
Journal of Nuclear Science and Technology, 57(1), p.9 - 23, 2020/01
Times Cited Count：1 Percentile：10.81(Nuclear Science & Technology)In sodium-cooled fast reactors (SFRs), it is necessary to remove the sodium remaining on spent fuel assemblies (FAs) before storing them in a spent fuel water pool (SFP) in order to minimize plant operating loads. A next-generation SFR in Japan has adopted an advanced dry cleaning process which consists of the following steps: argon gas blowing to remove the metallic residual sodium on the FA, moist argon gas blowing to deactivate the residual sodium, and direct storage in the SFP. This process increases economic competitiveness and reduces waste products. In this RD work, performance of the dry cleaning process has been investigated. This paper describes experimental and analytical work focusing on the amount of residual sodium remaining on FA components, for instance the handling head, the wrapper tube, the upper shielding, and the entrance nozzle which was conducted after investigation of residual sodium on fuel pin bundles as a part of series study of the cleaning process.
Segawa, Tomoomi; Yamamoto, Kazuya; Makino, Takayoshi; Iso, Hidetoshi; Kawaguchi, Koichi; Ishii, Katsunori; Sato, Hisato; Fukasawa, Tomonori*; Fukui, Kunihiro*
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.738 - 745, 2019/09
In the MOX fuel fabrication process, the dry grinding technology of mixed oxide pellets have been developed for the effective use of nuclear fuel materials. To develop a technology to control the particle size of dry recovered powder, the performance of the buhrstone mill and the collision plate type jet mill were studied using a simulated powder of particle size distribution about 500 m. We found that the particle size can be controlled at the range of about 250 m or less by both by adjusting the clearance between the grinding wheels of the buhrstone mill, and the clearance and elevation angle of the clarification zone of the collision plate type jet mill. And furthermore, the collision plate type jet mill is considered to be suitable for particle size control because the operating parameters of the classifier can be finely adjusted.
Kato, Atsushi; Nagai, Keiichi; Ara, Kuniaki; Otaka, Masahiko; Oka, Nobuki*; Tanaka, Masako*; Otani, Yuichi*; Ide, Akihiro*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
In a fuel handling system (FHS) of a sodium-cooled fast reactor, it is necessary to reduce residual sodium on a spent fuel subassembly before storing at a spent fuel water-pool (SFP) in order to minimize design loads. Although the wet cleaning process adopted on MONJU could eliminate almost all of residual sodium, a large amount of radioactive liquid waste occurs and it needs long duration of cleaning treatment and large plant commodities. On the other hand, Japan sodium-cooled fast reactor adopted an advanced dry cleaning system which consists of roughly blowing massive sodium on the fuel subassembly out by 300
C argon gas, inactivation of residual sodium to NaOH by moist argon gas and directly immersion into the SFP to achieve economic competitiveness and waste reduction. This paper reports current status of recent R&D activities to demonstrate a performance of the dry cleaning process in Japan which are for improvement of the cleaning performance and optimizing the FHS design.
Tsujimoto, Kazufumi; Sasa, Toshinobu; Maekawa, Fujio; Matsumura, Tatsuro; Hayashi, Hirokazu; Kurata, Masaki; Morita, Yasuji; Oigawa, Hiroyuki
Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.657 - 663, 2015/09
To continue the utilization of the nuclear fission energy, the management of the high-level radioactive waste is one of the most important issues to be solved. Partitioning and Transmutation technology of HLW is expected to be effective to mitigate the burden of the HLW disposal by reducing the radiological toxicity and heat generation. The Japan Atomic Energy Agency (JAEA) has been conducting the research and development on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. This paper overviews the recent progress and future R&D plan of the study on the ADS and related fuel cycle technology in JAEA.
Hayashi, Hirokazu; Nishi, Tsuyoshi; Takano, Masahide; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki
NEA/NSC/R(2015)2 (Internet), p.360 - 367, 2015/06
Uranium-free nitride fuel was chosen as the first candidate for transmutation of long-lived minor actinides (MA) using accelerator-driven system (ADS) under the double strata fuel cycle concept by Japan Atomic Energy Agency (JAEA). The advantages of nitride fuel are good thermal properties and large mutual solubility among actinide elements. A pyrochemical process is proposed as the first candidate for the reprocessing of the spent nitride fuel, because this technique has some advantages over aqueous process, such as the resistance to radiation damage, which is an important issue for the fuels containing large amounts of highly radioactive MA. This paper overviews the recent progress and future R&D plan of the study on the nitride fuel cycle technology in JAEA.
Okano, Fuminori; Ikeda, Yoshitaka; Sakasai, Akira; Hanada, Masaya
Dai-27-Kai Genshiryoku Shisetsu Dekomisshoningu Gijutsu Koza Tekisuto, p.73 - 102, 2014/10
The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the wielded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident.
Ogawa, Toru; Minato, Kazuo; Okamoto, Yoshihiro; Nishihara, Kenji
Journal of Nuclear Materials, 360(1), p.12 - 15, 2007/01
Times Cited Count：15 Percentile：70.17(Materials Science, Multidisciplinary)The actinide management has become a key issue in nuclear energy due to increasing proliferation concern and long-term environmental perception. The better way of waste management will be made by system symbiosis: a combination of light-water reactor and fast reactor and/or accelerator-driven transmutation system should be sought. The new recycling technology should be able to achieve good economy with smaller plants, which can process fuels from different types of reactors on a common technical basis. Pyroprocess with the use of molten salts is regarded as the strong candidate for such recycling technology. In JAEA, the first laboratory for the high temperature chemistry of transuranium elements, mainly Am and Cm, has been established. The fundamental data on the molten-salt chemistry of transuranium oxides and nitrides will be combined with the computer code for predicting the molten-salts electrolytic processes.
Shirai, Osamu*; Yamana, Hajimu*; Arai, Yasuo
Journal of Alloys and Compounds, 408-412, p.1267 - 1273, 2006/02
Times Cited Count：41 Percentile：84.4(Chemistry, Physical)no abstracts in English
Minato, Kazuo; Akabori, Mitsuo; Tsuboi, Takashi; Kurobane, Shiro; Hayashi, Hirokazu; Takano, Masahide; Otobe, Haruyoshi; Misumi, Masahiro*; Sakamoto, Takuya*; Kato, Isao*; et al.
JAERI-Tech 2005-059, 61 Pages, 2005/09
JAERI-Tech-2005-059.pdf:20.67MB
An experimental facility called the Module for TRU High Temperature Chemistry (TRU-HITEC) was installed in the Back-end Cycle Key Elements Research Facility (BECKY) of the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) for the basic studies of the behavior of the transuranium elements (TRU) in pyrochemical reprocessing and oxide fuels. TRU-HITEC consists of three alpha/gamma cells shielded by steel and polyethylene and a glove box shielded by leaded acrylic resin, where experimental apparatuses have been equipped and a high purity argon gas atmosphere is maintained. In the facility 10 g of 
Am as well as the other TRU of Np, Pu and Cm can be handled. This report summarizes the outline, structure, performance and interior apparatuses of the facility, and is the result of the joint research between the Japan Atomic Energy Research Institute and three electric power companies of Tokyo Electric Power Co., Tohoku Electric Power Co. and the Japan Atomic Power Co.
systemsOkamoto, Yoshihiro; Shiwaku, Hideaki; Yaita, Tsuyoshi; Suzuki, Shinichi; Minato, Kazuo; Tanida, Hajime*
Zeitschrift f
r Naturforschung, A, 59a(11), p.819 - 824, 2004/11
The local structure of molten CdCl and CdCl-KCl mixture was investigated by high-temperature XAFS technique. The nearest Cd-Cl distance decreases by melting. Similarly, the coordination number decreases from 6 to 4. It suggests that there is (CdCl)
tetrahedral complex ion in the melt. It is concluded that the local structure is kept also in the mixture melt, since the same XAFS result as the pure melt was obtained.
Okamoto, Yoshihiro; Minato, Kazuo
JAERI-Conf 2004-008, 228 Pages, 2004/04
JAERI-Conf-2004-008.pdf:17.05MB
Applications of molten salts technology to separation and synthesis of materials have a potential to give us a civilized life, for example aluminium refinement. Recently, much attention is given to the pyrochemical reprocessing of spent nuclear fuels in the molten salt research field. On the other hand, computer simulation technique is expected to play an important role for supporting experimental works and predicting unknown physical properties in the molten salts application studies. Research group for Actinides Science, Department of Materials Science, Japan Atomic Energy Research Institute(JAERI), together with Reprocessing and Recycle Technology Division, Atomic Energy Society of Japan, organized the 3rd Workshop on Molten Salts Technology and Computer Simulation at Tokai Research Establishment, JAERI on December 16, 2003. Many molten salts researchers in Japan participated in the workshop and many useful presentations and discussions were performed.
Watanabe, Takeshi*; Tsushima, Satoru*; Yamamoto, Ichiro*; Tomioka, Osamu; Meguro, Yoshihiro; Nakashima, Mikio; Wada, Ryutaro*; Nagase, Yoshiyuki*; Fukuzato, Ryuichi*
Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.363 - 366, 2004/00
Recovery of salts by supercritical fluid leaching (SFL) method using carbon dioxide was experimentally studied. It was confirmed that LiCl was recovered with a mixed fluid of carbon dioxide and methanol, and KCl and SrCl were recovered with a mixed fluid of carbon dioxide, methanol and crown ether. The influence of crown ether for KCl and SrCl extraction were found to increase in the order of 15-crown-5 (15C5) 
18-crown-6 (18C6) dicychlohexyl-18-crown-6 (DC18C6). It is expected that other salts can be recovered selectively with a mixed fluid of carbon dioxide, methanol and suitable crown ether.
