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Senzaki, Tatsuya; Arai, Yoichi; Yano, Kimihiko; Sato, Daisuke; Tada, Kohei; Ogi, Hiromichi*; Kawanobe, Takayuki*; Ono, Shimpei; Nakamura, Masahiro; Kitawaki, Shinichi; et al.
JAEA-Testing 2022-001, 28 Pages, 2022/05
JAEA-Testing-2022-001.pdf:2.33MB
In preparation for the decommissioning of Laboratory B of the Nuclear Fuel Cycle Engineering Laboratory, the nuclear fuel material that had been stored in the glove box for a long time was moved to the Chemical Processing Facility (CPF). This nuclear fuel material was stored with sealed by a polyvinyl chloride (PVC) bag in the storage. Since it was confirmed that the PVC bag swelled during storage, it seems that any gas was generated by radiolysis of the some components contained in the nuclear fuel material. In order to avoid breakage of the PVC bag and keep it safety for long time, we began the study on the stabilization treatment of the nuclear fuel material. First, in order to clarify the properties of nuclear fuel material, radioactivity analysis, component analysis, and thermal analysis were carried out. From the results of thermal analysis, the existence of organic matter was clarified. Then, ion exchange resin with similar thermal characteristics was selected and the thermal decomposition conditions were investigated. From the results of these analyzes and examinations, the conditions for thermal decomposition of the nuclear fuel material contained with organic matter was established. Performing a heat treatment of a small amount of nuclear fuel material in order to confirm the safety, after which the treatment amount was scaled up. It was confirmed by the weight change after the heat treatment that the nuclear fuel material contained with organic matter was completely decomposed.
Tada, Kohei; Kitawaki, Shinichi; Watanabe, So; Aihara, Haruka; Shibata, Atsuhiro; Nomura, Kazunori
Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 3 Pages, 2017/09
Radioactive liquid waste containing chloride ion (Cl) is generated by chemical analysis for process control of pyroprocessing. To realize discharging this liquid waste to the sea, it's necessary to carry out the process in order to separate Cl and recover U, Pu. This study carried out a combination of the AgCl precipitation method and extraction chromatography method to separate Cl and recover U, Pu. The result of precipitation test showed that U and Pu didn't occur the co-precipitation after the test. The result of solid phase extraction test showed that 95% of Pu was successfully recovered from the liquid waste. It was difficult to analyze 
radioactivity about U because the concentration of U is not enough. These results showed that these process has the feasibility of the discharging the liquid waste to the sea.
Watanabe, So; Nomura, Kazunori; Kitawaki, Shinichi; Shibata, Atsuhiro; Kofuji, Hirohide; Sano, Yuichi; Takeuchi, Masayuki
Procedia Chemistry, 21, p.101 - 108, 2016/12
Times Cited Count：14 Percentile：99.14(Chemistry, Inorganic & Nuclear)Aihara, Haruka; Kitawaki, Shinichi; Nomura, Kazunori; Taguchi, Katsuya
Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1076 - 1083, 2015/09
system using a combination of X-ray diffraction and differential thermal analysesNakayoshi, Akira; Kitawaki, Shinichi; Fukushima, Mineo; Murakami, Tsuyoshi*; Kurata, Masaki
Journal of Nuclear Materials, 441(1-3), p.468 - 472, 2013/10
Times Cited Count：14 Percentile：72.14(Materials Science, Multidisciplinary)Electrorefining is one of the main steps of pyroreprocessing where spent nuclear fuels are recycled. Electrorefining is conducted in a molten salt of LiCl-KCl eutectic (59:41 mol%) containing actinide chlorides (AnCl) at 773 K. In order to operate and maintain the electrorefiner, it is necessary to accumulate fundamental data on LiCl-KCl-AnCl salt such as the melting point. In this study, based on X-ray diffraction and differential thermal analysis, a partial phase diagram of (LiCl-KCl)eut.-UCl pseudo-binary system and partial phase diagram of LiCl-KCl-UCl system were developed, which UCl concentration was up to 20 mol%.
O by mechanochemical method with CCl
Nagai, Takayuki; Kitawaki, Shinichi; Sato, Nobuaki*
Materials Sciences and Applications, 4(7), p.419 - 431, 2013/07
For a chlorinating method at low temperature, the possibility of chlorination of NdO by a mechanochemical reaction with CCl was studied using a planetary ball mill. The mechanochemical experiments were conducted by changing the pot materials, milling time, molar ratio of CCl/NdO, and revolution speed. As the results of obtained products by X-ray diffractometry and Raman spectroscopy, it was confirmed that the chlorination to NdOCl from NdO with CCl was advanced at room temperature in a zirconia or tungsten pot with balls. We found that an extension of the milling time and an increase of the number of ball were effective to the chlorination to NdOCl and that tensile stress remained in the milled powder by using a planetary ball mill.
MoONagai, Takayuki; Sato, Nobuaki*; Kitawaki, Shinichi; Uehara, Akihiro*; Fujii, Toshiyuki*; Yamana, Hajimu*; Myochin, Munetaka
Journal of Nuclear Materials, 433(1-3), p.397 - 403, 2013/02
Times Cited Count：10 Percentile：61.16(Materials Science, Multidisciplinary)In order to examine easily synthetic conditions of uranyl molybdate, UOMoO, used for the reprocessing process study of spent nuclear oxide fuels in alkaline molybdate melts, the uranium molybdate compounds were produced from UO
powder and anhydrous MoO reagent. The results of having investigated them in solid state by using X-ray diffractometry and Raman spectrometry, it was confirmed that UOMoO could be synthesized by heating mixed powder of UO and MoO with stoichiometric mole ratio at 770
C for 4 h under air atmosphere. Moreover, adding this UOMoO into LiMoO-NaMoO eutectic melt, most of the dissolved uranium species in the melt were observed as hexa-valent uranyl ions by absorption spectrophotometry.
Kitawaki, Shinichi; Sakamura, Yoshiharu*
Denki Kagaku Oyobi Kogyo Butsuri Kagaku, 79(12), p.975 - 976, 2011/12
no abstracts in English
Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo; Sakamura, Yoshiharu*; Murakami, Tsuyoshi*; Akiyama, Naoyuki*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/09
In the FaCT project, the metal fuel cycle including metal fuel fast reactor and pyrochemical reprocessing has been being developed. JAEA and CRIEPI have continued a collaborative study on pyrochemical reprocessing. In the pyrochemical reprocessing, actinides in the spent fuels dissolve anodically in the LiCl-KCl, and U is collected selectively on a solid cathode, Pu and MA are recovered simultaneously in a liquid Cd cathode. In the previous electrorefining tests, at the anode Zr was allowed to dissolve into the electrolyte salt together with U, Pu and MA. The Zr co-dissolution may cause some problems. In this study, through the anode dissolution test of U-Pu-Zr alloy fuel, the controlling the dissolution of the Zr and the improvement of dissolution ratio of U, Pu were studied. The U-Pu alloy was prepared from MOX pellets by using the electrochemical reduction method. U-Pu-Zr ternary alloy was produced by alloying the obtained U-Pu alloy and prepared U-Zr alloy. U-Pu-Zr ternary alloy was immersed into electrolyte salt, and electrolysis test was carried out.
Murakami, Tsuyoshi*; Sakamura, Yoshiharu*; Akiyama, Naoyuki*; Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo
Journal of Nuclear Materials, 414(2), p.194 - 199, 2011/07
Times Cited Count：17 Percentile：77.43(Materials Science, Multidisciplinary)An electrorefining is one of the main steps of pyrochemical reprocessing of spent metallic fuels (U-Zr, U-Pu-Zr). The electrorefining is carried out dissolving a portion of Zr together with actinides to accomplish a high dissolution ratio of actinides. However, the electrorefining with Zr co-dissolution should bring some practical problems in the pyrochemical reprocessing. Therefore, electrorefining tests of non-irradiated U-Pu-Zr alloy were performed with minimizing the amount of Zr dissolved in LiCl-KCl-(U, Pu, Am)Cl melts at 773 K. The tests were performed both by potentiostatic electrolysis at -1.0 V (Ag
/Ag) that was more negative than the Zr dissolution potential and by galvanostatic electrolysis with a limited amount of Zr dissolution. The ICP-AES analysis of the anode residues confirmed that a high dissolution ratio of actinides (U; 
99.6%, Pu; 99.9%) was successfully demonstrated at both electrolyses.
Kofuji, Hirohide; Fukushima, Mineo; Kitawaki, Shinichi; Myochin, Munetaka; Kormilitsyn, M. V.*; Terai, Takayuki*
IOP Conference Series; Materials Science and Engineering, 9, p.012010_1 - 012010_8, 2010/05
Times Cited Count：0 Percentile：1.02(Chemistry, Inorganic & Nuclear)Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo; Koizumi, Tsutomu; Kurata, Masaki*; Yahagi, Noboru*
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1269 - 1273, 2009/09
JAEA is developing the pyroreprocessing by collaboration with CRIEPI. The test using U began in 2002, and the test using PuO and unirradiated MOX were ended in 2008. The reduction of UO pellets by using Li-reduction method, the electrowinning using reduced pellets, the separation of adhered salt with deposit by distillation, and the ingot formation of deposit were performed. As a result, 99% of the loaded U is recovered as metal ingot. The tests similar to U tests were performed by using PuO. As a result, Pu was successfully recovered with U metal. In the MOX test, the mass balance of Pu was maintained at 
100% with respect to the initial amount. We try to form the U-Pu-Zr alloy by using reduced MOX. After 2009, the process development that uses the alloy will be continued.
Kitawaki, Shinichi; Shinozaki, Tadahiro; Fukushima, Mineo; Usami, Tsuyoshi*; Yahagi, Noboru*; Kurata, Masaki*
Nuclear Technology, 162(2), p.118 - 123, 2008/05
Times Cited Count：18 Percentile：74.54(Nuclear Science & Technology)A series test of pyro-process was carried out to recover U-Pu alloy from MONJU MOX pellets. In the Li-reduction step, the reduction behavior of MOX was similar to that of UO. In the electrorefining step, the separation factor between U and Pu was 5.7 for the combination of reduced MOX anode and liquid cadmium cathode, which is almost comparable to the value in the previous studies. For the material balance, approximately 98% of U and 103% of Pu were detected in the electrodes or molten salt after the electrolysis with respect to the initial amounts in the anode or molten salt. Considering the analytical error of ICP-AES, these values are reasonable. The remained amount of U in the anode was slightly larger than that of Pu due to the re-oxidation. The U-Pu alloy ingot was successfully formed by distillation of Cd.
-UCl meltFukushima, Mineo; Nakayoshi, Akira; Kitawaki, Shinichi; Kurata, Masaki*; Yahagi, Noboru*
Proceedings of 3rd International ATALANTE Conference (ATALANTE 2008) (CD-ROM), 4 Pages, 2008/05
Products on solid cathodes recovered from metal pyrochemical processing were processed to obtain uranium ingot. Studies on process conditions for uranium forming, assay recovered uranium products and by-products and evaluation of mass balance were carried out. In this tests, it is confirmed that uranium ingots can be obtained with heating the products up to more than melting temperature of metal uranium under normal pressure because adhered salt cover the uranium not to oxidize it during uranium cohering. Volatilization of americium is very small under the condition of high temperature.
Nakayoshi, Akira; Kitawaki, Shinichi; Fukushima, Mineo; Kurata, Masaki*; Yahagi, Noboru*
Proceedings of International Symposium on EcoTopia Science 2007 (ISETS '07) (CD-ROM), p.1062 - 1066, 2007/11
Pyro-reprocessing is one of the promising reprocessing methods for recycling spent nuclear fuels generated from fast reactors. Comparing to the conventional aqueous-processes, following benefits are expected when introducing the pyro-reprocessing, such as reduction of environmental burden, enhancement of proliferation-resistant, enhancement of economical potential, efficient utilization of nuclear resources. The pyro-reprocessing will therefore become more attractive not only in developed countries regarding nuclear energy, but also in developing countries. As for reducing environmental burden, the most important subject is establishment of the nuclear fuel cycle, in which actinide elements are closed. Various kinds of intermediate waste which contains actinide elements are formed in the practical operation not only from the main steps of the pyro-reprocessing but also from related sub-streams.
Koyama, Tadafumi*; Hijikata, Takatoshi*; Usami, Tsuyoshi*; Inoue, Tadashi*; Kitawaki, Shinichi; Shinozaki, Tadahiro; Fukushima, Mineo; Myochin, Munetaka
Journal of Nuclear Science and Technology, 44(3), p.382 - 392, 2007/03
Times Cited Count：24 Percentile：82.59(Nuclear Science & Technology)Electrometallurgical pyroprocessing is a promising technology to realize actinide fuel cycle. Integrated experiments to demonstrate electrometallurgical pyroprocessing of plutonium oxide in continuous operation were carried out. In each test, 10 to 20 g of PuO was reacted with Li reductant to form metal product. The reduction products were charged in the anode basket of the electrorefiner with LiCl-KCl-UCl electrolyte. With using the anodes, deposition of uranium on the solid cathode was carried out, when PuCl concentration was low. After Pu/U ratio in salt electrolyte was increased enough, plutonium and uranium were recovered simultaneously on the liquid cadmium cathode. By heating up the deposits for distillation of the salt and the cadmium, U metal or Pu-U alloyed metal were obtained as residues in the crucible. It was first result to demonstrate the recovery of metal actinides in the continuous operation of pyroprocessing of oxide fuels.
Kitawaki, Shinichi; Shinozaki, Tadahiro; Fukushima, Mineo; Usami, Tsuyoshi*; Yahagi, Noboru*; Kurata, Masaki*
no journal, ,
no abstracts in English
Usami, Tsuyoshi*; Kurata, Masaki*; Yahagi, Noboru*; Kitawaki, Shinichi; Shinozaki, Tadahiro; Fukushima, Mineo
no journal, ,
no abstracts in English
Yahagi, Noboru*; Usami, Tsuyoshi*; Kurata, Masaki*; Kitawaki, Shinichi; Fukushima, Mineo; Shinozaki, Tadahiro
no journal, ,
no abstracts in English
Kurata, Masaki*; Usami, Tsuyoshi*; Yahagi, Noboru*; Kitawaki, Shinichi; Fukushima, Mineo; Shinozaki, Tadahiro
no journal, ,
no abstracts in English
