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Pu
Am)N solid solutionsNishi, Tsuyoshi; Takano, Masahide; Akabori, Mitsuo; Arai, Yasuo
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
The thermal conductivity of Zr-based minor actinide (MA) nitride solid solutions is important for designing subcritical cores in nitride-fueled ADS. However, there have been no experimental data on the thermal conductivities of Zr-based nitride solid solutions containing MA. In this study, the authors prepared sintered samples of (ZrPuAm)N (x=0.0, 0.58, 0.80) solid solutions. The thermal diffusivity and heat capacity of (ZrPuAm)N solid solutions were measured using a laser flash method and drop calorimetry, respectively. Thermal conductivities were determined from the measured thermal diffusivities, heat capacities and bulk densities over a temperature range of 473 to 1473 K. Moreover, in order to help to promote the design study of nitride-fueled ADS, the thermal conductivity of the (ZrPuAm)N solid solutions were fitted to an equation using the least squares method.
Nagaoka, Shinichi; Ishida, Michihiko; Kanamori, Sadamu; Hayashi, Shinichiro
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
The feasibility of applying PSA to nuclear fuel cycle facilities such as reprocessing plants has been also studied. We conducted a simplified risk assessment of each of the selected individual accident events and compared the assessment results for four accident categories (fire, explosion, criticality, and other accident events in which large amounts of radioactive materials are released).
Yanagisawa, Kazuaki
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
An economic scale of nuclear energy was evaluated as a total amount of sales of electricity. It was 43,018 million dollars, including a fuel cycle cost by 10,860 million dollars. Due to several casualty accidents, for example, an earth quake attacked to the Kashiwazaki-Kariwa Units in 2007, the economic scale of nuclear energy was decreasing. The indirect effect of nuclear energy linked with the green technology was effective to avoid global warming. Hypothetical trading of carbon dioxide emission might save 4,000 million dollars, that is 10 percent of the ordinary earnings.
Kitamura, Akira; Sasaki, Takayuki*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12
Thermodynamic interpretation on solubility of neptunium, technetium, selenium and palladium in nitrate and ammonium solutions was performed using the thermodynamic database developed by Japan Atomic Energy Agency (JAEA-TDB). We found that we had to pay special care of redox reactions of nitrogen in thermodynamic calculations. Although solubility data used in the present study were well interpreted using JAEA-TDB, we found that nitrate will affect redox conditions and further experimental studies were required to focus redox behavior of nitrogen.
Sudo, Katsuo; Takano, Tatsuo; Takeuchi, Kentaro; Kihara, Yoshiyuki; Kato, Masato
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
Japan Atomic Energy Agency has been contracted to advance the Fast Reactor Cycle Technology Development project. As one part of the project, a simplified MOX pellet fabrication method has been developed for fast reactor fuels. In previous reports, feasibility of a simplified MOX pellet fabrication method was confirmed through hot and cold laboratory-scale experiments. The die wall lubrication pressing technology was one of the important technologies included in the development of the simplified MOX pellet fabrication method. In the work described here, a pressing machine with a die wall lubrication system was developed, and MOX pellet fabrication experiments were carried out on the kilogram MOX scale.
Se in spent nuclear fuel sampleAsai, 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
Morihira, Masayuki; Mizusako, Fumiki*; Tsuboi, Yasushi*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
Cladding inner corrosion is one of the life controlling factors of FBR MOX fuels and it depends on the oxygen potential in a fuel element. Oxygen potential increases with extension of burn-up due to the cumulated excess oxygen during fission. The oxygen getter method is idea way to locate metal fragments in a fuel element as an excess oxygen absorber. Since almost nothing has been reported concerning the application of an oxygen getter in pellet type fuels, conceptual development of the oxygen getter for pellet type MOX fuel and a feasibility study were done. For getter material, titanium was mainly evaluated in this study except for compatibility tests carried out for titanium and zirconium. Concerning the location of getter material in a fuel element, the pellet-cladding gap and axial blanket region are potential options to avoid melting of titanium or obtaining a eutectic solution with MOX fuel. At the same time, an adequate temperature for oxidation as well as compatibilities with cladding material and fuel must be realized. Three options were proposed for titanium and their potentials were evaluated from this viewpoint. As a result, locating the titanium pellets in the upper axial blanket region of the fuel element was identified as the most promising option and it could provide the required low smear density titanium pellet.
Suzuki, Kazuyuki; Hatanaka, Akira; Samoto, Hirotaka; Suwa, Toshio; Tanaka, Kosuke; Tanaka, Yukiyoshi
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
The properties of the sludge in dissolver vessels from the reprocessing of ATR-MOX and ATR-UO
fuels were investigated on the pilot-plant scale at the Tokai Reprocessing Plant (TRP). This sludge is mainly composed of platinum-group elements, zircaloy fragments, and post-precipitates from the dissolver solution. The sludge deposited on the dissolver causes difficulties such as pipe clogging. The characteristics of the sludge collected from the dissolver vessels, which affect the reprocessing operation, were revealed through chemical composition analysis using ICP-AES, and XRD. It was confirmed that the major component of the sludge was zirconium molybdate, and no significant differences between ATR-MOX and ATR-UO fuels were observed in terms of the sludge compositions. In order to gain further understanding of the properties of the sludge, the distributions of Pu and other trace elements were EPMA.
Oigawa, Hiroyuki
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
When we explore the sustainable utilization of nuclear power, reasonable and environmentally preferable waste management is indispensable. The partitioning and transmutation technology has been studied aiming at reduction of the burden for disposal of high-level radioactive waste. As for the partitioning process of spent fuel, various innovative extractants and methods are being studied and proposed to separate minor actinide (MA) from lanthanide, and so on. As for the transmutation of long-lived nuclides, various types of system, such as MA loading to a fast reactor and dedicated transmutation of MA in an accelerator-driven system, are being studied and proposed, and respective types of MA-bearing fuel are being investigated. One of problems to proceed with research and development on this technology is in the difficulty to provide and handle a certain amount of MA. To overcome this point, international collaboration to make use of facilities and MA resources is desirable.
Yamanaka, Atsushi; Hashimoto, Kowa; Uchida, Toyomi; Shirato, Yoji; Isozaki, Toshihiko; Nakamura, Yoshinobu
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
The Tokai Reprocessing Plant (TRP) adopted the PUREX method in 1977 and has reprocessed spent nuclear fuel of 1140 tHM (tons of heavy metals) since then. The reprocessing equipment suffers from various corrosion phenomena because of high nitric acidity, solution ion concentrations, such as uranium, plutonium, and fission products, and temperature. Therefore, considering corrosion performance in such a severe environment, stainless steels, titanium steel, and so forth were employed as corrosion resistant materials. The severity of the corrosive environment depends on the nitric acid concentration and the temperature of the solution, and uranium in the solution reportedly does not significantly affect the corrosion of stainless steels and controls the corrosion rates of titanium steel. The TRP equipment that handles uranyl nitrate solution operates at a low nitric acid concentration and has not experienced corrosion problems until now. However, there is a report that corrosion rates of some stainless steels increase in proportion to rising uranium concentrations. The equipment that handles the uranyl nitrate solution in the TRP includes the evaporators, which concentrate uranyl nitrate to a maximum concentration of about 1000 gU/L (grams of uranium per liter), and the denitrator, where uranyl nitrate is converted to UO
powder at about 320
C. These equipments are therefore required to grasp the degree of the progress of corrosion to handle high-temperature and high-concentration uranyl nitrate. The evaluation of this equipment on the basis of thickness measurement confirmed only minor corrosion and indicated that the equipment would be fully adequate for future operation.
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; Shibata, Atsuhiro; Koma, Yoshikazu
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 4 Pages, 2011/12
To elucidate the behavior of fission products, two counter current experiments were carried out with a dissolver solution derived from irradiated fast neutron reactor "JOYO" core fuel, one was a single cycle flow sheet using double scrubbing solution with 9 and 1 mol/dm
HNO and the other used a Tc scrubbing solution with 10 mol/dm HNO. Among fission products, the decontamination behavior of Zr and Tc differed according to HNO concentration of the scrubbing solutions. The decontamination factor of Zr and Tc increase to 
76.8 and 7.52 with 10 mol/dm HNO in the Tc scrubbing solution. Other fission products such as Cs was well decontaminated and its DF resulted in 10
.
Takemoto, Noriyuki; Izumo, Hironobu; Hori, Naohiko; Ishitsuka, Etsuo; Ishihara, Masahiro
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 4 Pages, 2011/12
Establishment of a world network is proposed to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008 every year, and a common understanding of world network has begun to be shared. The ISMTR-5 will be held on USA in 2012, and the ISMTR-6 will be held on Argentine in 2013.
Sano, Yuichi; Kaji, Naoya; Shibata, Atsuhiro; Takeuchi, Masayuki; Washiya, Tadahiro
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12
-tetradodecyldiglycolamide with the aid of a process simulation code, PARC-MAMorita, Yasuji; Tsubata, Yasuhiro; Sasaki, Yuji; Kimura, Takaumi
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
A separation process for trivalent actinides and rare earths from high-level liquid waste (HLLW) by extraction with -tetradodecyldiglycolamide (TDdDGA) is being developed by performing counter-current continuous extraction tests using mixer-settler and by calculating element behavior with a process simulation code. The continuous extraction test with a simulated HLLW showed that Am was recovered with 0.1M TDdDGA-n-dodecane in a high yield. Calculations of process simulation for the continuous extraction test and for optimized process were performed by PARC-MA which was developed by JAEA. The results of the calculation by PARC-MA agreed well with the results of the continuous extraction tests. In a optimized process condition, La (and a part of the lighter lanthanides) can be transferred to the raffinate keeping the high extraction yield of Am. TDdDGA extraction system has an ability to treat HLLW of high element concentration.
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.
Kunii, Katsuhiko; Gonda, Mayuki; Ikeda, Kiyoshi; Nagaya, Shun; Itabashi, Keizo; Nakajima, Hidemitsu; Mineo, Yukinobu
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
The Library of the Japan Atomic Energy Agency (JAEA) has developed two Japanese literature materials of bibliographic tools based on the International Nuclear Information System (INIS) of the IAEA which contains over 3.3 million records of 127 countries and 24 international organizations. These materials have been elaborated by appropriately designating Japanese terminology of nuclear field corresponding with English terminology or vice versa. One is "Transliterated Japanese journal title list" and the other is "INIS Thesaurus in Japanese". While the former is served as a reference that enables users to access articles of Japanese journals better matching their needs, the latter is served as a dictionary to bridge the gap on nuclear field terminologies between over 30,000 English terms and Japanese terms which correspond with those in a semantic manner. The application of those materials to the INIS's full text collection over 280,000 of technical reports, proceedings etc. as an archive is helpful for enhancement of human resource development. The authors describe the effectiveness of those INIS-based materials with bibliographic references of Fukushima Daiichi NPS accident.
Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
An advanced LWR with hard neutron spectrum, named FLWR, aims at efficient and flexible utilization of nuclear resources by evolving its fuel assembly design under the same core configuration, mainly corresponding to available fuel cycle technologies and related infrastructures. The paper summarizes an evolution process of the FLWR fuel assembly design toward a sustainable fuel cycle by dividing the reactor operation into three stages, that is, the one based on the current LWR MOX fuel cycle infrastructure such as reprocessing of UO spent fuel and fabrication of MOX fuel, the one for transitioning from the LWR fuel cycle to the FR fuel cycle, and the one based on the FR fuel cycle infrastructures such as MOX spent fuel reprocessing.
Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12
Waste management of the fast breeder reactor (FBR) fuel cycle with and without partitioning and transmutation (P&T) was investigated by focusing thermal constraints in storage and disposal facilities. The result showed that transmutation of minor actinides (MAs) is essentially effective to reduce the waste emplacement area in repository, and a combination of P&T can provide a compact disposal with a smaller emplacement area than the conventional repository design by two orders of magnitude. Cost analysis revealed that the cost for storage and disposal is comparable among the conventional light water reactor, FBR without P&T, and FBR only with MA-transmutation. The cost of disposal for FBR fuel cycle with P&T is significantly reduced by an order of magnitude from the others, while that of storage does not increase.
Delage, F.*; Arai, Yasuo; Belin, R.*; Chen, X.*; D'Agata, E.*; Hania, R.*; Klaassen, F.*; Maschek, W.*; Oigawa, Hiroyuki; Ottaviani, J. P.*; et al.
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
The European FP-6 integrated project EUROTRANS was devoted to management of high-level wastes from nuclear power plants, focusing on MA transmutation in ADS. The object of the project was the assessment of the design and feasibility of an industrial ADS prototype dedicated to MA transmutation. JAEA joined in the project as one of partners. This paper summarizes the ADS fuel development carried out in this project. As for the oxide fuel, a primary candidate, the results of design study, performance in normal operation, safety analysis, irradiation tests and out-of-pile property measurements are described. As for the nitride fuel, an alternative of oxide fuel, the results of irradiation tests and out-of-pile property measurements, and the progress of pyrochemical process for spent fuel treatment are described.