Asai, Shiho; Toshimitsu, Masaaki; Hanzawa, Yukiko; Suzuki, Hideya; Shinohara, Nobuo; Inagawa, Jun; Okumura, Keisuke; Hotoku, Shinobu; Kimura, Takaumi; Suzuki, Kensuke*; et al.
Journal of Nuclear Science and Technology, 50(6), p.556 - 562, 2013/06
The Sn content in a spent nuclear fuel solution was determined by ICP-MS for its inventory estimation in high-level radioactive waste. An irradiated UO fuel was used as a sample to evaluate the reliability of the methodology. Prior to the measurement, Sn was separated from Te, which causes major isobaric interference in the determination of Sn content, along with highly radioactive coexisting elements using an anion-exchange column. The absence of counts attributed to Te in the Sn-containing effluent indicates that Te was completely removed. After washing, Sn retained on the column was readily eluted with 1 M HNO. The isotope ratios of Sn were successfully determined and showed good agreement with those obtained through ORIGEN2 calculations. The results reported in this paper are the first experimental values of Sn content in the spent nuclear fuel solution originating in spent nuclear fuel irradiated at a nuclear power plant in Japan.
Asai, Shiho; Hanzawa, Yukiko; Suzuki, Hideya; Toshimitsu, Masaaki; Okumura, Keisuke; Shinohara, Nobuo; Kimura, Takaumi; Inagawa, Jun; Suzuki, Kensuke*; Kaneko, Satoru*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
Asai, Shiho; Okumura, Keisuke; Hanzawa, Yukiko; Suzuki, Hideya; Toshimitsu, Masaaki; Inagawa, Jun; Kimura, Takaumi; Kaneko, Satoru*; Suzuki, Kensuke*
Proceedings of 14th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2011) (CD-ROM), p.1437 - 1442, 2011/09
Okumura, Keisuke; Asai, Shiho; Hanzawa, Yukiko; Okamoto, Tsutomu; Suzuki, Hideya; Toshimitsu, Masaaki; Inagawa, Jun; Kimura, Takaumi; Suzuki, Kensuke*; Kaneko, Satoru*
Proceedings of 14th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2011) (CD-ROM), p.1443 - 1450, 2011/09
Inventory estimation of long-lived fission products (LLFPs) in high-level radioactive wastes (HLW) from spent nuclear fuels of light water reactors is important for a safety assessment of their disposal. In order to develop an inventory estimation method of difficult-to-measure LLFPs (Se-79, Tc-99, Sn-126, and Cs-135), a parametric study was carried out by using a sophisticated burnup calculation code and data. In the parametric study, fuel specifications and irradiation conditions are changed in the conceivable range. The considered parameters are fuel assembly types (PWR / BWR), U-235 enrichment, moderator temperature, void fraction, power density, and so on. From the calculated results, we clarify the burnup characteristics of the target LLFPs and their possible ranges of generations. Finally, candidates of the key nuclide are proposed for the scaling factor method of HLW.
Asai, Shiho; Hanzawa, Yukiko; Okumura, Keisuke; Shinohara, Nobuo; Inagawa, Jun; Hotoku, Shinobu; Suzuki, Kensuke*; Kaneko, Satoru*
Journal of Nuclear Science and Technology, 48(5), p.851 - 854, 2011/05
Asai, Shiho; Hanzawa, Yukiko; Okumura, Keisuke; Suzuki, Hideya; Toshimitsu, Masaaki; Shinohara, Nobuo; Kaneko, Satoru*; Suzuki, Kensuke*
Proceedings of 13th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2010) (CD-ROM), p.261 - 264, 2010/10
Ishikawa, Masumi*; Kaneko, Satoru*; Kitayama, Kazumi*; Ishiguro, Katsuhiko*; Ueda, Hiroyoshi*; Wakasugi, Keiichiro*; Shinohara, Nobuo; Okumura, Keisuke; Chino, Masamichi; Moriya Noriyasu*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 8(4), p.304 - 312, 2009/12
Since quality control issues for vitrified waste are defined mainly with the focus on the transport and storage of the waste rather than the long-term safety of geological disposal, they do not cover inventories of long-lived nuclides which are of most interest in the safety assessment of geological disposal. Therefore we suggest a flow chart for assessment of inventories of long-lived nuclides in the vitrified waste focusing on measured value. We started a programme to examine the applicability as well as to improve reliability of nuclide generation/decay code and nuclear data library using liquid waste from spent fuel with clear irradiation history. To solve the issue of quality control for vitrified waste, comprehensive study is needed in aspects not only of geological disposal field but also of operation of nuclear power plant, reprocessing of spent fuel and vitrification of liquid waste. This study is a pioneering study to integrate them.
Otobe, Haruyoshi; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo
Journal of the American Ceramic Society, 92(1), p.174 - 178, 2009/01
The oxygen potentials of the oxygen-deficient fluorite-type oxide AmPuO were measured by the electrochemical method with using a zirconia solid-electrolyte. The coulomb titration has been made for the sample at 1333 K over 0.02 0.25. The oxygen potentials were -93.63 and -440.18 kJmol for = 0.021 and 0.25 at 1333 K, respectively. The temperature dependence of the oxygen potentials was also measured between 1000 and 1333 K over the range of 0.02 0.243. The temperature dependence was almost linear over the and temperature ranges concerned.
Otobe, Haruyoshi; Akabori, Mitsuo; Minato, Kazuo
Journal of the American Ceramic Society, 91(6), p.1981 - 1985, 2008/06
The oxygen potentials of AmO were measured in the range of 0.01 to 0.5 and the temperature range of 1000 to 1333 K by the electromotive force (EMF) method. The oxygen potentials at 1333 K were -19.83 kJ/mol for =0.01 and -319.1 kJ/mol for =0.485, which were higher than those of CeO by approximately 200 kJ/mol for the corresponding values. From the dependence of the oxygen potentials on and temperature, a tentative phase diagram of Am-O system was proposed, which suggested the presence of the intermediate phases of AmO and AmO in the Am-O system.
Hayashi, Hirokazu; Takano, Masahide; Akabori, Mitsuo; Minato, Kazuo
Journal of Alloys and Compounds, 456(1-2), p.243 - 246, 2008/05
Americium trichloride was synthesized by the reaction of americium nitride with cadmium chloride at 600-660 K in a dynamic vacuum. The product was hexagonal AmCl, of which lattice parameters were determined to be = 0.7390 and = 0.4215 nm. The results indicate that high purity AmCl samples, in which the oxychloride was not found, were prepared without the use of corrosive reagents. The reaction of the nitrides with cadmium chloride is suitable for synthesis of high purity actinide and lanthanide chlorides.
Amaya, Masaki; Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 45(5), p.423 - 431, 2008/05
Pulse irradiation simulating reactivity initiated accident (RIA) condition was conducted for the test rod prepared from the BWR fuel rod with a burnup of 56 GWd/t irradiated in a commercial reactor, and fission gas release during the pulse irradiation was investigated based on the result of rod-puncture test and electron-probe-microanalysis of fuel pellet. The local xenon concentration of pulse-irradiated pellet decreased compared with that of base-irradiated pellet in the relative radius of 0 to approximately 0.8. The decrease corresponds to a fractional fission gas release of approximately 11%, and this value was comparable with the rod-puncture test result. Considering the microstructural change in fuel pellet and the amount of retained gas in grain boundary, it is likely that the fission gas release during pulse irradiation was affected by the grain boundary separation which occurred in the mid-radius rather than the peripheral region of fuel pellet during pulse irradiation.
Otobe, Haruyoshi; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo
Proceedings of 3rd International ATALANTE Conference (ATALANTE 2008) (CD-ROM), 4 Pages, 2008/05
The oxygen potentials of cubic zirconia containing Pu and americia have been measured by the electromotive force (EMF) method with a zirconia solid-electrolyte. The oxygen potentials of these oxides were reviewed. The phase relations, microstructure, equilibrium state of these oxides were discussed, referring to the isothermal curve of the oxygen potentials. It was found that the oxygen potentials are sensitive to the phase transitions. Moreover, the study on the microstructure around an element incorporated in oxide fuels as well as its phase is necessary to predict the effect of the element on the oxygen potentials of oxide fuels.
Nishi, Tsuyoshi; Takano, Masahide; Ito, Akinori; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo; Numata, Masami
Journal of Nuclear Materials, 373(1-3), p.295 - 298, 2008/02
The thermal diffusivity of americium oxide was determined in the temperature range from 299 to 1473 K by a laser flash method. The thermal diffusivity of AmO decreased with increasing temperature. The thermal conductivity of AmO was estimated from the measured thermal diffusivity, the specific heat capacity and the bulk density. It was found that the thermal conductivity of AmO decreased with increasing temperature over the temperature range investigated. It was also found that the decrease in O/Am ratio during the thermal diffusivity measurements under vacuum resulted in a slight decrease in thermal conductivity of AmO.
Sugikawa, Susumu; Nakazaki, Masato; Kimura, Akihiro; kida, Takashi*; Kihara, Takehiro*; Akabori, Mitsuo; Minato, Kazuo; Suda, Kazuhiro*; Chikazawa, Takahiro*
Nippon 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.
Hayashi, Hirokazu; Akabori, Mitsuo; Minato, Kazuo; Mizuguchi, Koji*; Kawabe, Akihiro*; Fujita, Reiko*
Denki Kagaku Oyobi Kogyo Butsuri Kagaku, 75(7), p.528 - 534, 2007/07
The simulation code for the pyrochemical processing of spent nuclear fuels was developed to analyze experimental data, to predict experimental results, and to propose adequate conditions and processes. The Simulation code for Pyrochemical Reprocessing (SPR) is based on calculations of chemical equilibrium and electrochemical reactions. The code also includes the calculations of the current-potential distribution between the electrodes. Some calculations were made to simulate the experimental results on the electro-codeposition process of UO and PuO. The phenomena of the redox reactions between Pu and Pu ions and those between Fe and Fe ions were theoretically analyzed; these redox reactions cause the low current efficiency in the electro-codeposition process. The calculated current-potential distribution around the cathode corresponds to the observed distribution of the oxide deposited on the cathode.
Yamagishi, Hideshi; Kakuta, Tsunemi*
Kyodo Kenkyu Seika Gaiyo, Shoraigata Genshiryoku Hatsuden Gijutsu No Kodoka Ni Kansuru Kenkyu, p.197 - 211, 2006/06
no abstracts in English
Hayashi, Hirokazu; Akabori, Mitsuo; Minato, Kazuo
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 3 Pages, 2005/10
For a basis of the future nuclear cycle, it is very important to understand and control the behavior of TRU (Np, Pu, Am, Cm) in the nuclear fuel cycle. Experimental study of pyrochemical process of fuels containing TRU requires the facility having not only shielding for -ray and neutron but also ability to keep a high purity inert gas atmosphere; because minor actinide chlorides can easily react with oxygen or water vapor in an atmosphere. The module for TRU high temperature chemistry (TRU-HITEC) had been installed to study the basic properties of TRU in the pyrochemical processes. In the present work, the behavior of Am in pyrochemical process was investigated by electrochemical methods.
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
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.
The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi
JAERI-Tech 2004-049, 44 Pages, 2004/06
At the Kashiwazaki-Kariwa Nuclear Power Station Unit-1, indications of cracks were identified in a weld joint portion of the primary loop recirculation piping. To investigate the cause of cracks, TEPCO conducted a material examination on the specimen including the cracks. The present investigation was carried out to ensure transparency of the examination by providing JAERI's own evaluation report as a third party organization. The following findings were made; (1) A crack was observed near the weld region. (2) Intergranular cracking was observed at almost whole fracture surface. (3) Transgranular cracking was observed at the crack opening region. Increases of hardness by cold work were observed and the crack was initiated near the region where hardness value showed the highest. (4) Content of Cr was very slightly depleted in the vicinity of grain boundary. Based on the above results with the presence of tensile residual stress near the crack generated by welding process and dissolved oxygen contents in cooling water, the observed cracks were concluded to be stress corrosion cracking.
The Working Team for Examination Operation of Samples From Core Shroud at Fukushima Dai-ni Unit-3
JAERI-Tech 2004-044, 92 Pages, 2004/05
The present examination has been performed with the objective to ensure the transparency of the examination as the third-party organization by providing technical basis for identifying the causes of cracking through examination of the sample taken from the cracked region of outer H6a welding portion of the core shroud at Fukushima Dai-ni Nuclear Power Station Unit-3, which was a part of sample stored in the Nippon Nuclear Fuel Development Co., Ltd. in the examination of Tokyo Electric Power Company in 2001. The present examination of the sample was conducted at the post irradiation examination facilities of JAERI. The following findings were obtained from the result of the present examination. (1)Three cracks were observed at the portion 3 to 9mm apart from the weld metal and the maximum depth was about 8mm. (2)Intergranular cracking was observed in almost whole fracture surface. The transgranular cracking was partially observed within the depth of about 300m from the surface. (3)Hardening layer over Hv400 at its maximum was found from the surface to the depth of about 500m. Based on the examination results concerning presence of tensile residual stress by welding and relatively high dissolved oxygen contents in core coolant, it is concluded that the cracks were mainly initiated in the hardening layer by transgranular stress corrosion cracking and propagated along the grain boundaries.