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Takeuchi, Masayuki; Takata, Takeshi; Saito, Keita*; Chikazawa, Takahiro*
Progress in Nuclear Science and Technology (Internet), 7, p.135 - 141, 2025/03
Control of insoluble sludge from fuel dissolution process is one of the important issues to secure the safety in reprocessing plant operation. In order to achieve the higher sludge recovery in the clarification stage, new clarification system which is an integrated technology of centrifugal device and filter unit has been developed for spent MOX fuel reprocessing. In this study, over-all clarification performance of this integrated system was evaluated on engineering scale. As results, total sludge recovery rate of more than 99.5 % was achieved in all test conditions by this system from the engineering test and this new clarification system showed excellent clarification performance. The sludge recovery rate of centrifugal device was influenced by test conditions and was in the range from 87 to 98 %. This study showed this technology is one of the promising clarification systems to improve the performance of sludge recovery greatly.
Nagata, Masanobu; Chikazawa, Takahiro*; Akahori, Kuniaki*; Kitamura, Akira; Tachi, Yukio
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 23(1), p.31 - 54, 2016/06
Although spent nuclear fuel is planned to be disposed after reprocessing and vitrification of high-level radioactive waste (HLW), feasibility study on direct disposal of spent nuclear fuel (SF) has been started as an alternative option to flexibly apply change of future energy situation in Japan. Radionuclide inventories and their release behavior after breaching spent fuel container should be assessed to confirm safety of the SF disposal. However, these detailed studies have not been performed in Japan. Therefore, we investigated some foreign safety assessment reports on direct disposal of spent nuclear fuel by focusing on the source term of the fast release of radionuclides (i.e. instant release fraction; IRF) for the purpose of contributing to the safety assessment of Japanese SF disposal system. As a result of comparison between the safety assessment reports in foreign countries, although some fundamental data have been referred to the reports in common, the final source term dataset (IRF) was seen differences between countries in the result of taking into account the national circumstances (Reactor type and burnup, etc.). We also found the difference of assignment of uncertainties among the investigated reports; a report selected pessimistic values and another report selected mean values and their deviations. It is expected that these findings are useful as fundamental information for determination of the release rates for the safety assessment of Japanese SF disposal system.
Kitamura, Akira; Chikazawa, Takahiro*; Akahori, Kuniaki*; Tachi, Yukio
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 23(1), p.55 - 72, 2016/06
The Japanese geological disposal program has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter "direct disposal of SF") as an alternative management option other reprocessing followed by vitrification and geological disposal of high-level radioactive waste. We conducted literature survey of dissolution rate of SF matrix and constructing materials (e.g. zircaloy cladding and control rods) selected in safety assessment reports for direct disposal of SF in Europe and United States. We also investigated basis of release rate determination and assignment of uncertainties in the safety assessment reports. Furthermore, we summarized major conclusions proposed by some European projects governed by European Commission. It was found that determined release rates are fairly similar to each other due to use of similar literature data in all countries of interest. It was also found that the determined release rates were including conservativeness because it was difficult to assign uncertainties quantitatively. It is expected that these findings are useful as fundamental information for determination of the release rates for the safety assessment of Japanese SF disposal system.
Takeuchi, Masayuki; Yano, Kimihiko; Shibata, Atsuhiro; Sambommatsu, Yuji*; Nakamura, Kazuhito*; Chikazawa, Takahiro*; Hirasawa, Izumi*
Journal of Nuclear Science and Technology, 53(4), p.521 - 528, 2016/04
Times Cited Count：3 Percentile：23.79(Nuclear Science & Technology)Chikazawa, Yoshitaka; Yoshiuji, Takahiro*
Nuclear Engineering and Design, 289, p.1 - 7, 2015/08
Times Cited Count：10 Percentile：57.53(Nuclear Science & Technology)A phased array acoustic leak detection system for sodium heated steam generator has been proposed. The major advantage of the new system is it could provide information of acoustic source direction. An acoustic source of a sodium-water reaction is supposed to be localized while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore the new system could separate the target leak source from steam generator background noise. In the previous study, the methodology was proposed and basic performance was confirmed by numerical analysis. However, in the numerical analysis, acoustic transportation through the SG tube bundle was not modeled. In the present study, performance the proposed system has been confirmed in water experiments with mockup tube bundles.
Ikeda, Yasuhisa*; Kawasaki, Takeshi*; Harada, Masayuki*; Nogami, Masanobu*; Kawata, Yoshihisa*; Kim, S.-Y.*; Morita, Yasuji; Chikazawa, Takahiro*; Someya, Hiroshi*; Kikuchi, Toshiaki*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
An advanced reprocessing system for spent FBR fuels based on two precipitation processes using pyrrolidone derivatives as precipitants has been developed. Experimental results of precipitation behavior of U, Pu and other elements, the heat- and radiation-resistance of precipitants, the thermal decomposition properties of precipitates showed that N-n-butyl-2-pyrrolidone and N-neopentyl-2-pyrrolidone are the appropriate precipitants for the first and second precipitation steps, respectively. From the engineering investigation, We confirmed that the precipitation and the filtration can be done efficiently using the engineering scale equipment and that the fuel pellets are directly prepared by the calcination of the precipitates. On the basis of these results, we evaluated that the proposed system is expected to be one of candidates of the future reprocessing systems for spent FBR fuels.
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.
Nakahara, Masaumi; Nomura, Kazunori; Washiya, Tadahiro; Chikazawa, Takahiro*; Hirasawa, Izumi*
Journal of Nuclear Science and Technology, 48(3), p.322 - 329, 2011/03
Times Cited Count：8 Percentile：53.37(Nuclear Science & Technology)The purification behavior of uranyl nitrate hexahydrate was investigated to evaluate the decontamination performance of liquid and solid impurities using a dissolver solution of mixed oxide fuel in batch experiments by the sweating and the melt filtration processes. Liquid impurities such as Eu were effectively removed by the sweating method, but solid impurities such as Pu, Cs and Ba were affected a little in the batch experiments. On the other hand, the decontamination factor of Ba increased with 0.45 and 5.0 
m filters in the melt filtration process. Although the decontamination factors of Pu and Cs did not change with 5.0 m filter, it increased approximately two-fold with 0.45 m filter. The particle size of Cs
Pu(NO
)
is a little small and might pass through the 5.0 m filter in the melt filtration process.
Nakahara, Masaumi; Nomura, Kazunori; Washiya, Tadahiro; Chikazawa, Takahiro*; Hirasawa, Izumi*
Radiochimica Acta, 98(6), p.315 - 320, 2010/06
Times Cited Count：7 Percentile：42.29(Chemistry, Inorganic & Nuclear)In order to examine the decontamination behavior of the Pu and fission products in the U crystallization process, experiments were carried out using mixed oxide fuel dissolver solution. It is confirmed that Eu was decontaminated by washing the uranyl nitrate hexahydrate crystals. However, the decontamination factors of Ba and Cs were low because they precipitated as Ba(NO) and CsPu(NO) after the U crystallization, respectively. The decontamination factor of Cs tends to decrease with increasing HNO and Pu concentrations in the mother liquor because of the generation of CsPu(NO).
Washiya, Tadahiro; Tayama, Toshimitsu; Nakamura, Kazuhito*; Yano, Kimihiko; Shibata, Atsuhiro; Nomura, Kazunori; Chikazawa, Takahiro*; Nagata, Masanobu*; Kikuchi, Toshiaki*
Journal of Power and Energy Systems (Internet), 4(1), p.191 - 201, 2010/02
Japan Atomic Energy Agency (JAEA) and Mitsubishi Materials Corporation (MMC) are developing the crystallization process for elemental technology of FBR fuel reprocessing. The uranium (U) crystallization process is a key technology for New Extraction System for TRU Recovery (NEXT) process that was evaluated as the most promising process for future FBR reprocessing. We had developed an innovative crystallizer and fabricated an engineering-scale crystallizer and have carried out continuous operation test to investigate the stability of the equipment at steady and non-steady state conditions by using depleted uranium. As for simulating typical failure events in the crystallizer, crystal accumulation and crystal blockage were occurred intentionally, and monitoring method and resume procedure were tried and selected in this work.
Nakahara, Masaumi; Nomura, Kazunori; Washiya, Tadahiro; Chikazawa, Takahiro*; Hirasawa, Izumi*
Journal of Alloys and Compounds, 489(2), p.659 - 662, 2010/01
Times Cited Count：3 Percentile：18.18(Chemistry, Physical)Characterization of Cs and Pu(IV) nitrate complex was examined to separation from the UNH crystal. This complex is obtained as a precipitate by mixing dissolver solution of MOX fuel and CsNO solution, which was identified to dicesium tetravalent plutonium hexanitrate, CsPu(NO) by concentration analysis and XRD. The precipitate has a tendency to be generated at high HNO condition. Thermal analysis shows that the precipitate is stable below 245 
C, and a weight loss of about 10.29% is observed at 245 C. This result corresponds to the decomposition of CsPu(NO) to CsPuO(NO)
. According to these properties, the UNH crystal can be melted at 60 C to 100 C, and separated from the Pu(IV) and Cs complex by a filtration. A new technology of crystal purification method aimed at higher decontamination of UNH crystal in the U crystallization process is proposed.
Shibata, Atsuhiro; Yano, Kimihiko; Kamiya, Masayoshi; Nakamura, Kazuhito; Washiya, Tadahiro; Chikazawa, Takahiro*; Kikuchi, Toshiaki*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 8(3), p.245 - 253, 2009/09
Behavior of Cs in U crystallization process of advanced aqueous reprocessing system was investigated with simulated dissolver solution. Beaker-scale U crystallization experiments were carried out with some simulated dissolver solutions. The results show that possibility of generation of CsNO,CsUO (NO) or Cs-FP complex salt is small. Precipitation experiments of Cs-U(IV) complex salts were also carried out with nitrate solution of U(IV) and Cs. It was found that Cs-U(IV) complex salt was precipitated in higher acidity than 5 mol/dm
. It is suggested that Cs-Pu(IV) precipitates can be generated in the U crystallization process, under specific solution condition.
Shibata, Atsuhiro; Kaji, Naoya; Nakahara, Masaumi; Yano, Kimihiko; Tayama, Toshimitsu; Nakamura, Kazuhito; Washiya, Tadahiro; Myochin, Munetaka; Chikazawa, Takahiro*; Kikuchi, Toshiaki*
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.151 - 157, 2009/09
As a part of FaCT project, Japan Atomic Energy Agency has been developing a U crystallization process for advanced aqueous reprocessing technology in collaboration with Mitsubishi Materials Corporation. We have carried out experimental studies and obtained fundamental data. Continuous operation tests were also carried out by an engineering-scale crystallizer to confirm productivity of the equipment and to investigate non-steady state conditions. The requirements for the U crystallization process in the FaCT project could be achieved except DF of Cs. More detail investigation is under way to settle the process condition without Pu-Cs double salt formation.
Yano, Kimihiko; Nakahara, Masaumi; Nakamura, Masahiro; Shibata, Atsuhiro; Nomura, Kazunori; Nakamura, Kazuhito*; Tayama, Toshimitsu; Washiya, Tadahiro; Chikazawa, Takahiro*; Kikuchi, Toshiaki*; et al.
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.143 - 150, 2009/09
Washiya, Tadahiro; Tayama, Toshimitsu; Nakamura, Kazuhito; Yano, Kimihiko; Shibata, Atsuhiro; Nomura, Kazunori; Chikazawa, Takahiro*; Nagata, Masanobu*; Kikuchi, Toshiaki*
Proceedings of the 17th International Conference on Nuclear Engineering (ICONE-17), p.83 - 88, 2009/07
Japan Atomic Energy Agency (JAEA) and Mitsubishi Materials Corporation(MMC) are developing the crystallization process for FBR fuel reprocessing. The uranium (U) crystallization process is a key technology for NEXT(New Extraction System for TRU Recovery) process that was evaluated as the most promising process for future FBR reprocessing. We fabricated an engineering-scale crystallizer and continuous operation test was carried out to investigate steady and non-steady state conditions. Crystal accumulation and blockage phenomena were confirmed in this test. And monitoring method and resume procedure were considered. As the test results, no significant event was observed in the steady state test. And in the non-steady state test, process fluctuation was detected by the monitor of the screw torque and liquid level in the crystallizer, and all failure events can be recovered by appropriate resumed procedures.
Chikazawa, Takahiro*; Kikuchi, Toshiaki*; Shibata, Atsuhiro; Koyama, Tomozo; Homma, Shunji*
Journal of Nuclear Science and Technology, 45(6), p.582 - 587, 2008/06
Times Cited Count：21 Percentile：76.09(Nuclear Science & Technology)Batch crystallization of uranyl nitrate is carried out in order to obtain fundamental data required for the development of reprocessing involving crystallization. Particular attention is paid to the development of a method for predicting the concentrations of uranium and nitric acid in the mother liquor and the amount of uranyl nitrate crystals produced. Initial concentrations of uranyl nitrate and nitric acid are 500-600 g/dm
and 4-6 mol/dm, respectively, corresponding to the condition of a dissolver solution of spent fuel. Steady-state mass balance equations including the correlation equation for the equilibrium solubility of uranium nitrate are applied to the prediction. The calculated concentrations of uranium and nitric acid are in close agreement with the experimental ones. The recovery of uranium is accurately predicted by the calculated concentrations, with an error of less than 5%.
Homma, Shunji*; Ishii, Junichi; Kikuchi, Toshiaki*; Chikazawa, Takahiro*; Shibata, Atsuhiro; Koyama, Tomozo; Koga, Jiro*; Matsumoto, Shiro*
Journal of Nuclear Science and Technology, 45(6), p.510 - 517, 2008/06
Times Cited Count：13 Percentile：61.58(Nuclear Science & Technology)U-Pu co-crystallization reprocessing system is proposed for LWR fuels and its flowsheet study is carried out. This reprocessing system is based on the experimental evidence indicating that hexavalent plutonium is co-crystallized with uranyl nitrate. The system consists of five steps: dissolution of spent fuel, Pu oxidation, U-Pu co-crystallization, dissolution of the crystals, and U crystallization. The system does not require organic solvent, expecting the enhancement of safety and cost-effectiveness. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step to recover almost entire amount of U and Pu. The appropriate recycle ratio is determined for LWR fuels, such that the satisfactory decontamination is achieved. The mother liquor from the U re-crystallization step contains U and Pu. The flowsheet study shows that the constant ratio of Pu to U in the mother liquor can be controlled even though the composition of the spent fuel is different.
Sugikawa, Susumu; Nakazaki, Masato; Kimura, Akihiro; Kida, Takashi*; Kihara, Takehiro*; Akabori, Mitsuo; Minato, Kazuo; Suda, Kazuhiro*; Chikazawa, Takahiro*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(4), p.476 - 483, 2007/12
A one-step simple extraction chromatography method using TODGA (
-tetraoctyl-diglycolamide) adsorbent column has been developed to separate the americium from plutonium-solvent extraction raffinate. The raffinate contained Am(
620 mg/
), Np(107 mg/), Ag(2000 mg/), Fe(290 mg/), Cr(38 mg/), Ni(52 mg/) and trace of TBP. Small-scale and scale-up tests for separation of americium and conversion to americium oxide were carried out in NUCEF. Efforts were made to increase yield and purity of americium. The americium was separated with 83-92% yields and 97-98% purities by small-scale tests and 85-95% yields and 98-99% purities by scale-up tests. The yields for conversion of americium nitrate solution to americium oxide were 89-100% by small-scale tests and 85-96 % by scale-up tests. Approximately 1.8 gram americium oxide was recovered from 6 litres of the raffinate and supplied for the research on the high-temperature chemistry of TRU.
Oyama, Koichi; Nomura, Kazunori; Washiya, Tadahiro; Tayama, Toshimitsu; Yano, Kimihiko; Shibata, Atsuhiro; Komaki, Jun; Chikazawa, Takahiro*; Kikuchi, Toshiaki*
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1461 - 1466, 2007/09
Japan Atomic Energy Agency (JAEA) has been developing the crystallization process technology in cooperation with Mitsubishi Materials Corporation, Saitama University and Waseda University. We have carried out experimental studies with uranium, MOX and spent fuel dissolved solution, and flowsheet analysis was researched. Crystal refinement study has been started to get more purified crystal. In association with these studies, an innovative continuous crystallizer and its system was developed to ensure high process performance. From the design study, an annular type continuous crystallizer was selected as the most promising design, and performance was confirmed by small-scale test and engineering scale demonstration at uranium crystallization conditions. In this paper, the research and development of crystallization process are described.
Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Ikeda, Yasuhisa*; Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*
Journal of Nuclear Science and Technology, 44(3), p.354 - 360, 2007/03
Times Cited Count：16 Percentile：69.88(Nuclear Science & Technology)Precipitation behavior of Pu and other transuranium elements with N-cyclohexyl-2-pyrrolidone (NCP) has been examined to develop a simple reprocessing based only on precipitation method. From HNO solutions containing only Pu, both Pu(VI) and Pu(IV) were precipitated with NCP, but they required more NCP than in the U(VI) precipitation. Selective U(VI) precipitation from HNO solution containing U(VI) and Pu(IV) was achieved by stirring the solution for sufficient time after addition of NCP with ratio of [NCP]/[U]=1.4. Addition of an enough amount of NCP to U(VI)-Pu(VI) or U(VI)-Pu(IV) solutions gave a quantitative precipitation of both U and Pu. Neither Am(III) nor Np(V) was precipitated in the selective U precipitation and the simultaneous U-Pu precipitation. These results demonstrate the feasibility of the reprocessing by precipitation with NCP.
