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Journal Articles

Evaluation of source term parameters for spent fuel disposal in foreign countries, 2; Dissolution rates of spent fuel matrices and construction materials for fuel assemblies

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.

Journal Articles

Evaluation of source term parameters for spent fuel disposal in foreign countries, 1; Instant release fraction from spent fuel matrices and composition materials for fuel assemblies

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.

Journal Articles

Purification of uranium products in crystallization system for nuclear fuel reprocessing

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:2 Percentile:25.63(Nuclear Science & Technology)

Journal Articles

Water experiment on phased array acoustic leak detection system for sodium-heated steam generator

Chikazawa, Yoshitaka; Yoshiuji, Takahiro*

Nuclear Engineering and Design, 289, p.1 - 7, 2015/08

 Times Cited Count:6 Percentile:54.79(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.

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

Development of advanced reprocessing system based on precipitation method using pyrrolidone derivatives as precipitants; Overall evaluation of system

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.

Journal Articles

Removal of liquid and solid impurities from uranyl nitrate hexahydrate crystalline particles in crystal purification process

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:5 Percentile:42.62(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 $$mu$$m filters in the melt filtration process. Although the decontamination factors of Pu and Cs did not change with 5.0 $$mu$$m filter, it increased approximately two-fold with 0.45 $$mu$$m filter. The particle size of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ is a little small and might pass through the 5.0 $$mu$$m filter in the melt filtration process.

Journal Articles

Influence of nitric acid and plutonium concentrations in dissolver solution of mixed oxide fuel on decontamination factors for uranyl nitrate hexahydrate crystal

Nakahara, Masaumi; Nomura, Kazunori; Washiya, Tadahiro; Chikazawa, Takahiro*; Hirasawa, Izumi*

Radiochimica Acta, 98(6), p.315 - 320, 2010/06

 Times Cited Count:6 Percentile:43.83(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$$_{3}$$)$$_{2}$$ and Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ after the U crystallization, respectively. The decontamination factor of Cs tends to decrease with increasing HNO$$_{3}$$ and Pu concentrations in the mother liquor because of the generation of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$.

Journal Articles

Continuous operation test at engineering scale uranium crystallizer

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.

Journal Articles

Preparation and characterization of dicesium tetravalent plutonium hexanitrate

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:2 Percentile:21.24(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$$_{3}$$ solution, which was identified to dicesium tetravalent plutonium hexanitrate, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ by concentration analysis and XRD. The precipitate has a tendency to be generated at high HNO$$_{3}$$ condition. Thermal analysis shows that the precipitate is stable below 245 $$^{circ}$$C, and a weight loss of about 10.29% is observed at 245 $$^{circ}$$C. This result corresponds to the decomposition of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ to Cs$$_{2}$$PuO$$_{2}$$(NO$$_{3}$$)$$_{4}$$. According to these properties, the UNH crystal can be melted at 60 $$^{circ}$$C to 100 $$^{circ}$$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.

Journal Articles

Experimental study on behavior of Cs in uranium crystallization of advanced aqueous reprocessing system with simulated dissolver solution

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$$_{3}$$,Cs$$_{2}$$UO$$_{2}$$ (NO$$_{3}$$)$$_{4}$$ 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$$^{-3}$$. It is suggested that Cs-Pu(IV) precipitates can be generated in the U crystallization process, under specific solution condition.

Journal Articles

Research and development of crystal purification for product of uranium crystallization process

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

Journal Articles

Current status on research and development of uranium crystallization system in advanced aqueous reprocessing of FaCT project

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.

Journal Articles

Batch crystallization of uranyl nitrate

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:17 Percentile:76.24(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$$^{3}$$ and 4-6 mol/dm$$^{3}$$, 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%.

Journal Articles

Flowsheet study of U-Pu Co-crystallization reprocessing system

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:11 Percentile:62.49(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.

Journal Articles

Separation of americium from plutonium-solvent extraction raffinate and conversion to americium oxide

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 (${it N,N,N,N'}$-tetraoctyl-diglycolamide) adsorbent column has been developed to separate the americium from plutonium-solvent extraction raffinate. The raffinate contained Am($$sim$$620 mg/$$l$$), Np($$sim$$107 mg/$$l$$), Ag($$sim$$2000 mg/$$l$$), Fe($$sim$$290 mg/$$l$$), Cr($$sim$$38 mg/$$l$$), Ni($$sim$$52 mg/$$l$$) 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.

Journal Articles

Development of uranium crystallization system in "NEXT" reprocessing process

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.

Journal Articles

A Study on precipitation behavior of plutonium and other transuranium elements with N-cyclohexyl-2-pyrrolidone for development of a simple reprocessing process

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:13 Percentile:69.08(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$$_{3}$$ 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$$_{3}$$ 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.

Journal Articles

Development of crystallizer for advanced aqueous reprocessing process

Washiya, Tadahiro; Kikuchi, Toshiaki*; Shibata, Atsuhiro; Chikazawa, Takahiro*; Homma, Shunji*

Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 7 Pages, 2006/07

The crystallization is an advanced and remarkable technology in the future reprocessing process, which requires safety and cost advantages. Japan Atomic Energy Agency (JAEA), Mitsubishi Materials Corporation and Saitama University have been developing an annular-type continuous crystallizer. This paper mainly discussed about this crystallizer design and its development. JAEA has considered following two application processes of the crystallization technology. One is a uranium crystallization process, which applied before the solvent extraction process to recover excessive uranium from dissolver solution and reducing the throughput in the later extraction process. In this process, highly concentrated dissolver solution (about 500g-HM/L) is fed to this crystallizer, and only uranium is crystallized. Another is a plutonium co-crystallization process, which consists of two crystallization steps and excludes extraction process, and thus it's expected to reduce the waste generation and to improve operation safety. In this process, plutonium is co-crystallized with uranium in the first step and separated from residual solution, then the crystals are dissolved into nitric acid solution and excessive uranium is crystallized in the second step. This residual solution is recycled to fuel dissolution process, thus it contributes to reduce nitric acid quantity consumption. For both crystallization processes, same crystallizer design can be applied; we have developed a continuous crystallization system to establish high process throughput and optimizing of the crystallization processes. In the design study of the crystallizer, an annular-type was selected as the most promising design. The fundamental data was obtained by scale-down test device with uranium conditions, and an engineering scale crystallizer was fabricated to confirm the system performance in engineering scale.

Journal Articles

Development of a simple reprocessing process using selective precipitant for uranyl ions; Precipitation behaviors of plutonium and other transuranium elements

Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Ikeda, Yasuhisa*; Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

N-cyclohexyl-2-pyrrolidone (NCP) can selectively precipitate U(VI) ions in aqueous nitric acid solutions. Utilizing this property, we have been developing a simple reprocessing process of spent nuclear fuel based only on precipitation method. In the first precipitation step, only U is separated by precipitation in a yield of about 70%, and in the second precipitation step both U and Pu are recovered and separated from fission products (FP) and other transuranium elements (TRU). In JAERI, precipitation behaviors of Pu and other TRU were examined experimentally, and the results showed the feasibility of the process establishement.

61 (Records 1-20 displayed on this page)