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Journal Articles

Characterization and storage of radioactive zeolite waste

Yamagishi, Isao; Nagaishi, Ryuji; Kato, Chiaki; Morita, Keisuke; Terada, Atsuhiko; Kamiji, Yu; Hino, Ryutaro; Sato, Hiroyuki; Nishihara, Kenji; Tsubata, Yasuhiro; et al.

Journal of Nuclear Science and Technology, 51(7-8), p.1044 - 1053, 2014/07

 Times Cited Count:16 Percentile:77.12(Nuclear Science & Technology)

For safe storage of zeolite wastes generated by treatment of radioactive saline water at the Fukushima Daiichi Nuclear Power Station, properties of the Herschelite adsorbent were studied and its adsorption vessel was evaluated for hydrogen production and corrosion. Hydrogen production depends on its water level and dissolved species because hydrogen is oxidized by radicals in water. It is possible to evaluate hydrogen production rate in Herschelite submerged in seawater or pure water by taking into account of the depth effect of the water. The reference vessel of decay heat 504 W with or without residual pure water was evaluated for the hydrogen concentration by thermal hydraulic analysis using obtained fundamental properties. Maximum hydrogen concentration was below the lower explosive limit (4 %). The steady-state corrosion potential of a stainless steel 316L increased with absorbed dose rate but its increase was repressed by the presence of Herschelite. At 750 Gy/h and $$<$$60$$^{circ}$$C which were values evaluated at the bottom of the vessel of 504 W, the localized corrosion of SUS316L contacted with Herschelite would not immediately occur under 20,000 ppm of Cl$$^{-}$$ concentration.

Journal Articles

Safe storage of zeolite adsorbents used for treatment of accident-generated water at Fukushima Daiichi Power Station

Yamagishi, Isao; Nagaishi, Ryuji; Terada, Atsuhiko; Kamiji, Yu; Kato, Chiaki; Morita, Keisuke; Nishihara, Kenji; Tsubata, Yasuhiro; Ji, W.*; Fukushima, Hisashi*; et al.

IAEA-CN-211 (Internet), 5 Pages, 2013/01

Since the accident at Fukushima Daiichi Nuclear Power Station, a large amount of radioactive contaminated water has been generated to cool damaged reactor cores. Adsorption of cesium with zeolite-like media was employed for treatment of this contaminated saline water. As spent zeolite media are highly radioactive, their safe storage is a pressing issue. Japan Atomic Energy Agency has extensively conducted R&D on the management of secondary wastes produced by the operation of the treatment system. Subjects on the safe storage of spent zeolites include the analysis of their characteristics and the evaluation of effectiveness of the present safety measures in consideration of decay heat emission and hydrogen generation by water radiolysis as well as durability of vessels exposed to saline. Preliminary results obtained are described in the present paper.

Journal Articles

Burning of MOX fuels in LWRs; Fuel history effects on thermal properties of hull and end piece wastes and the repository performance

Hirano, Fumio; Sato, Seichi*; Kozaki, Tamotsu*; Inagaki, Yaohiro*; Iwasaki, Tomohiko*; Oe, Toshiaki*; Kato, Kazuyuki*; Kitayama, Kazumi*; Nagasaki, Shinya*; Niibori, Yuichi*

Journal of Nuclear Science and Technology, 49(3), p.310 - 319, 2012/03


 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the fuel histories including the burn-up of UO$$_{2}$$ spent fuels, the cooling period before reprocessing, the storage period of fresh MOX fuels. The heat generation rates of hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO$$_{2}$$ spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80$$^{circ}$$C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 GWd-MOX needs to be limited to a value of 0.7 to 1.6, which is significantly lower than the value of 4.0 for 45 GWd-UO$$_{2}$$.

Journal Articles

J-ACTINET activities of training and education for actinide science research

Minato, Kazuo; Konashi, Kenji*; Yamana, Hajimu*; Yamanaka, Shinsuke*; Nagasaki, Shinya*; Ikeda, Yasuhisa*; Sato, Seichi*; Arita, Yuji*; Idemitsu, Kazuya*; Koyama, Tadafumi*

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Actinide science research is indispensable to maintain sustainable development of innovative nuclear technology. For actinide science research, special facilities with containment and radiation shields are needed to handle actinide materials. The number of facilities for actinide science research has been decreased, especially in universities, due to the high maintenance cost. J-ACTINET was established in 2008 to promote and facilitate actinide science research and to foster many of young scientists and engineers in actinide science. The research program was carried out, through which young researchers were expected to learn how to make experiments with advanced experimental tools and to broaden their horizons. The summer schools and computational science school were held to provide students and young researchers with the opportunities to come into contact with actinide science research. The overseas dispatch program was also carried out.

Journal Articles

Effect of exchangeable cations on apparent diffusion of Ca$$^{2+}$$ ions in Na- and Ca-montmorillonite mixtures

Kozaki, Tamotsu*; Sawaguchi, Takuma; Fujishima, Atsushi; Sato, Seichi*

Physics and Chemistry of the Earth, 35(6-8), p.254 - 258, 2010/00

 Times Cited Count:18 Percentile:54.77(Geosciences, Multidisciplinary)

Compacted Na-bentonite, of which the major mineral is montmorillonite, is a candidate buffer material for the geological disposal of high-level radioactive waste. A potential alteration of the bentonite in a repository is the partial replacement of the exchangeable cations of Na$$^{+}$$ with Ca$$^{2+}$$. The Ca$$^{2+}$$ cations could be released from cementitious materials and diffuse into the buffer material in the repository. In this study, to evaluate the alteration that could reduce the performance of the bentonite buffer, the apparent diffusion coefficients of HTO and Ca$$^{2+}$$ ions were determined from non-steady, one-dimensional diffusion experiments using Na- and Ca-montmorillonite mixtures with different ionic equivalent fractions of Ca$$^{2+}$$ ions. The apparent diffusion coefficient of HTO at a dry density of 1.0 Mg m$$^{-3}$$ slightly increased with an increase in the ionic equivalent fraction of Ca$$^{2+}$$ ions. However, the apparent diffusion coefficient of Ca$$^{2+}$$ and the activation energy for diffusion at the same dry density were independent of the ionic equivalent fraction of Ca$$^{2+}$$ ions. These findings suggest that unlike HTO, which can be postulated to diffuse mainly in pore water, Ca$$^{2+}$$ ion diffusion could occur predominantly in interlayer spaces.

Journal Articles

Flexible fuel cycle R&D for the smooth FBR deployment

Fukasawa, Tetsuo*; Yamashita, Junichi*; Hoshino, Kuniyoshi*; Sasahira, Akira*; Inoue, Tadashi*; Minato, Kazuo; Sato, Seichi*

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Transition period from light water reactors (LWR) to fast breeder reactors (FBR) is quite important to achieve the future FBR cycle system. The transition scenarios were carefully studied and the Flexible Fuel Cycle Initiative (FFCI) was proposed in this study. FFCI carries out about 90% uranium (U) removal from LWR spent fuels (SF) at first and then recovers plutonium/uranium (Pu/U) from the remaining SF named "recycle material"(RM) (about 40% U, 15% Pu and 45% other nuclides) for FBR fresh fuel fabrication according to the FBR deployment status. The FFCI has some merits compared with ordinary system that carries out full reprocessing of LWR SF, that is volume reduction of LWR SF by its conversion to RM (proliferation resistant material), and storage and supply of high Pu density RM according to FBR deployment rate changes.

Journal Articles

Uranium recovery in LWR reprocessing and plutonium/residual uranium conditioning in FBR reprocessing for the transition from LWR to FBR

Fukasawa, Tetsuo*; Yamashita, Junichi*; Hoshino, Kuniyoshi*; Sasahira, Akira*; Inoue, Tadashi*; Minato, Kazuo; Sato, Seichi*

Proceedings of 3rd International ATALANTE Conference (ATALANTE 2008) (CD-ROM), 7 Pages, 2008/05

In order to flexibly manage the transition period from LWR to FBR, the authors investigated the transition scenario and proposed the Flexible Fuel Cycle Initiative (FFCI). In FFCI, LWR spent fuel reprocessing only carries out the removal of about 90% uranium that will be purified and utilized in LWR after re-enrichment. The residual material (40% U, 15% Pu and 45% other nuclides) is transferred to temporary storage and/or FBR spent fuel reprocessing to recover Pu/U followed by FBR fresh fuel fabrication depending on the FBR introduction status. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR spent fuels, that is smaller LWR reprocessing facility, spent LWR fuel reduction, storage and supply of high proliferation resistant and high Pu density material that can flexibly respond to FBR introduction rate changes. The Pu balance was calculated under several cases, which revealed that the FFCI could supply enough Pu to FBR in any cases.

Journal Articles

Development of a performance analysis code for vibro-packed MOX fuels

Ishii, Tetsuya; Nemoto, Junichi*; Asaka, Takeo; Sato, Seichi*; Mayorshin, A.*; Shishalov, O.*; Kryukov, F.*

Journal of Nuclear Science and Technology, 45(4), p.263 - 273, 2008/04

 Times Cited Count:2 Percentile:17.09(Nuclear Science & Technology)

In order to develop a vibro-packed MOX fuel performance analysis code, thermochemical and mechanical properties of the vibro-packed fuels were incorporated into a pellet type fuel performance analysis code CEDAR. Calculations were made by the developed code on a vibro-packed MOX fuel pin irradiated at BN-600 in Russia. Since the calculated results agreed well with the behaviors obtained from the experimental data, it can be concluded that the code was well modeled and qualitatively validated.

Journal Articles

Kinetic behavior of water as migration media in compacted montmorillonite using H$$_{2}$$$$^{18}$$O and applying electric potential gradient

Tanaka, Shingo*; Noda, Natsuko*; Higashihara, Tomohiro*; Sato, Seichi*; Kozaki, Tamotsu*; Sato, Haruo; Hatanaka, Koichiro

Physics and Chemistry of the Earth, 33(Suppl.1), p.S163 - S168, 2008/00

In order to identify mass transport pathway in compacted bentonite, water transport behavior in compacted montmorillonite which is the major clay mineral constituent of the bentonite was studied. Back-to-back diffusion and electro-osmosis experiments for H$$_{2}$$O were carried out at montmorillonite densities of 1.0, 1.2 and 1.4 Mg/m$$^{3}$$ using H$$_{2}$$$$^{18}$$O as a tracer. Apparent diffusivities from the diffusion experiments and advection velosities and hydraulic dispersities from the electro-osmosis experiments were determined. The mass transport pathways were discussed by comparing with concentration profiles and peak positions of He, Na and Cl which were reported in the past. The hydraulic dispersities decreased in the order of He, H$$_{2}$$O, Cl and Na, and these differences were considered to be due to that transport pathway depended on species and hydraulic dispersity for each species also depended on transport pathway.

Journal Articles

Characterization of homoionic Fe$$^{2+}$$-type montmorillonite; Potential chemical species of iron contaminant

Kozai, Naofumi; Inada, Koichi*; Adachi, Yoshifusa*; Kawamura, Sachi*; Kashimoto, Yusuke*; Kozaki, Tamotsu*; Sato, Seichi*; Onuki, Toshihiko; Sakai, Takuro; Sato, Takahiro; et al.

Journal of Solid State Chemistry, 180(8), p.2279 - 2289, 2007/08

 Times Cited Count:14 Percentile:48.46(Chemistry, Inorganic & Nuclear)

Fe$$^{2+}$$-montmorillonite with Fe$$^{2+}$$ ions occupying cation exchange sites is an ideal transformation product in bentonite buffer material. We previously prepared a Fe$$^{2+}$$-montmorillonite sample using a FeCl$$_{2}$$ solution under an inert gas condition. This study attempted to determine the potential contaminant iron chemical species in the sample. It was found that a small amount of Cl$$^{-}$$ ions remained dispersed throughout the sample. The Cl$$^{-}$$ ion retention may be due to the adsorption of FeCl$$^{+}$$ in the initial FeCl$$_{2}$$ solution and the subsequent containment of the Cl$$^{-}$$ ions that are dissociated from the FeCl$$^{+}$$ during excess salt removal treatment. The latter may be explained by the slow release of the remaining Cl$$^{-}$$ ions from the collapsed interlayer of the montmorillonite or the transformation of a minor fraction of the remaining FeCl$$^{+}$$ to iron (III) hydroxide chloride complexes having low solubility.

Journal Articles

Lattice thermal expansions of (Dy,Zr)N solid solutions

Takano, Masahide; Tagami, Susumu; Minato, Kazuo; Kozaki, Tamotsu*; Sato, Seichi*

Journal of Alloys and Compounds, 439(1-2), p.215 - 220, 2007/07

 Times Cited Count:15 Percentile:63.36(Chemistry, Physical)

ZrN is a possible candidate for the diluent material of the nitride fuel containing minor actinides. In the present study, the lattice thermal expansions of ZrN, DyN and (Dy,Zr)N solid solutions were measured by high temperature XRD, as the analogous fuel material. The average linear thermal expansion coefficients of these nitrides increased from 7.86$$times$$10$$^{-6}$$ to 9.54$$times$$10$$^{-6}$$ K$$^{-1}$$ with increasing Dy fraction. On the analogy with the results on the composition dependence of the thermal expansion coefficients, the preferable effect of ZrN as the diluent material is suggested.

Journal Articles

A New method for Fe(II)-montmorillonite preparation using Fe(II)-nitrilotriacetate complex

Manjanna, J.*; Kozaki, Tamotsu*; Kozai, Naofumi; Sato, Seichi*

Journal of Nuclear Science and Technology, 44(7), p.929 - 932, 2007/07

 Times Cited Count:18 Percentile:75.54(Nuclear Science & Technology)

This study developed a new preparation method of Fe(II)-montmorillonite. In this method, first, a Fe(II)-NTA complex solution was prepared by dissolving iron oxides with NTA and a reducing agent. Next, montmorillonite was immersed in this Fe(II)-NTA solution to allow montmorillonite adsorb Fe$$^{2+}$$ ions. The advantages of this method are (1) inert-gas atmosphere is not needed for the entire preparation process and (2) no anions such as Cl$$^{-}$$ ions, which form ion pairs with Fe$$^{2+}$$ ions, are not included in all the chemicals used.

JAEA Reports

Announced document collection of the 1st Information Exchange Meeting on Radioactive Waste Disposal Research Network (Joint research)

Nakayama, Shinichi; Nagasaki, Shinya*; Inagaki, Yaohiro*; Oe, Toshiaki*; Sasaki, Takayuki*; Sato, Seichi*; Sato, Tsutomu*; Tanaka, Satoru*; Tochiyama, Osamu*; Nagao, Seiya*; et al.

JAEA-Conf 2007-003, 120 Pages, 2007/03


The 1st information exchange meeting on Radioactive Waste Disposal Research Network was held in Nuclear Science Research Institute of Japan Atomic Energy Agency on August 4, 2006. Radioactive Waste Disposal Research Network was established by under Interorganization Atomic Energy Research Program of Japan Atomic Energy Agency, and the objective is to bring both research infrastructures and human expertise in Japan to an adequate performance level, thereby contributing to the development of the fundamental research area in the field of radioactive waste disposal. This lecture material is a collection of research presentations and discussions during the information exchange meeting.

JAEA Reports

Effects of Repository Environment on Diffusion Behavior of Radionuclides in Buffer Materials (II)

Kozaki, Tamotsu*; Sato, Seichi*

JNC TJ8400 2004-022, 39 Pages, 2005/02


Compacted bentonite is considered as a candidate buffer material in the geological disposal of high-level radioactive waste. An ssential function of the compacted bentonite is to retard the transport of radionuclides from waste forms to the surrounding host rock after degradation of an overpack. Therefore, diffusion behavior of radionuclides in the compacted bentonite is one of the important issues to be studied for the safety assessment of the geological disposal. In this study, through-diffusion experiments were performed for sodium ions in compacted Na-montmorillonite, since sodium ions are the major exchangeable cations of the Na-montmorillonite which may affect the diffusion behaviors of radionuclides. Cumulate fluxes of$$^{22}$$Na ions diffusing into the montmorillonite (inlet) and out of the montmorillonite (outlet) were monitored as a function of time at the diffusion experiments under different diffusion temperature. The best-fitted parameters of effective diffusion coefficient and capacity factor to the cumulate fluxes of outlet were obtained by the analysis with or without the consideration of isotopic dilution of$$^{22}$$Na in the diffusion system. However, the parameters could not fit to the cumulate flux of inlet. This suggests that there must be still unknown diffusion process in the diffusion system.

JAEA Reports

Effects of Repository Environment on Diffusion Behavior of Radionuclides in Buffer Materials

Kozaki, Tamotsu*; Sato, Seichi*

JNC TJ8400 2003-075, 34 Pages, 2004/03


Compacted bentonite is considered as a candidate buffer material in the geological disposal of high-level radioactive waste. An important function of the compacted bentonite is to retard the transport of radionuclides from waste forms to the surrounding host rock after degradation of an overpack. Therefore, diffusion behavior of radionuclides in the compacted bentonite has been extensively studied by many researchers for the performance assessments of the geological disposal. However, diffusion mechanism of radionuclides in the bentonite cannot be fully understood, and most experimental data have been obtained at room temperature for the bentonite saturated with low salinity water, which would disagree often with real repository conditions. In this study, therefore, apparent diffusion coefficients were determined at various diffusion temperatures for chloride ions in Na-montmorillonite samples saturated with NaCl solution of high salinity. Activation energies for the apparent diffusion were also obtained from the temperature dependents of the diffusion coefficients at different salinity. As the salinity increased, the apparent diffusion coefficients of chloride ions in montmorillonite were found to increase slightly. On the other hand, the activation energies for the chloride diffusion were found to be almost constant (approximately 12 kJ mol$$^{-1}$$) and less than that in free water (17.4 kJ mol$$^{-1}$$). Effects of salinity on diffusion behavior of radionuclides in montmorillonite werediscussed from the viewpoints of microstructure of montmorillonite and distribution of ions in the montmorillonite. As a result, the diffusion behavior of sodium ions could be explained by the changes of the predominant diffusion process among pore water diffusion, surface diffusion, and interlayer diffusion that could be caused by the increase of salinity.

Journal Articles

Characterization of Fe-montmorillonite; A Simulant of buffer materials accommodating overpack corrosion product

Kozai, Naofumi; Adachi, Yoshifusa*; Kawamura, Sachi*; Inada, Koichi*; Kozaki, Tamotsu*; Sato, Seichi*; Ohashi, Hiroshi*; Onuki, Toshihiko; Bamba, Tsunetaka

Journal of Nuclear Science and Technology, 38(12), p.1141 - 1143, 2001/12

This study briefly reports characteristics of Fe-montmorillonite. Fe-montmorillonite was used as a simulant of buffer material in which corrosion products of carbon steel overpack, Fe$$^{2+}$$, were diffused. We have found that this clay retains Se(VI) for which natural montmorillonite, such as Na+-type and Ca$$^{2+}$$-type, has little retentivity.

Journal Articles

Apparent diffusion coefficients and chemical species of neptunium(V) in compacted na-montmorillonite

Kozai, Naofumi; Inada, Koichi*; Kozaki, Tamotsu*; Sato, Seichi*; Ohashi, Hiroshi*; Bamba, Tsunetaka

Journal of Contaminant Hydrology, 47(2-4), p.149 - 158, 2001/02

 Times Cited Count:15 Percentile:42.11(Environmental Sciences)

no abstracts in English

Journal Articles

Observation of Microstructures of Compacted Bentonite by Microfocus X-Ray Computerized Tomography (Micro-CT)

Suzuki, Satoru; Sato, Seichi*; *; Kozai, Naofumi*

Journal of Nuclear Science and Technology, 38(8), p.697 - 699, 2001/00

 Times Cited Count:22 Percentile:81.55(Nuclear Science & Technology)


Journal Articles

Concept and applicability of sorption distribution coefficient in the radionuclide transport model

*; Takasu, Aki*; *; Kimura, Hideo; Sato, Seichi*; Nagasaki, Shinya*; Nakayama, Shinichi; Niibori, Yuichi*; *; Mitsugashira, Toshiaki*; et al.

Genshiryoku Bakkuendo Kenkyu, 5(1), p.3 - 19, 1998/08

no abstracts in English

52 (Records 1-20 displayed on this page)