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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.
Seki, Masayuki; Kihara, Yoshiyuki; Kaito, Takeji; Tsukada, Tatsuya*; Motoki, Kazuhiko*; Hirako, Kazuhito*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
Oxide Dispersion Strengthened (ODS) steel has been developed as an advanced fuel cladding tube for sodium cooled fast reactors in Japan Atomic Energy Agency (JAEA) to attain the target burn up of 150 GWd/t in the bundle average because of its excellent swelling resistance and high mechanical strength in high temperature. If conventional TIG welding is applied to the ODS welding, it is difficult to obtain necessary mechanical strength at the weld zone because of the formation of porosity. It is formed by the argon bubbles which initially dissolve in the matrix and grow up under the high temperature during welding. Therefore JAEA has been conducted the development of pressurized resistance welding (PRW) technology for ODS cladding tube, which is one of the solid state welding methods. This paper describes in the development of PRW technology, an ultrasonic test method for detecting weld defects, the result of machine strength measurement examination in weld part and the result of fuel pin irradiation examination using nuclear reactor.
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.
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.
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.
Takano, Tatsuo; Sudo, Katsuo; Takeuchi, Kentaro; Kihara, Yoshiyuki; Kato, Masato
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12
Development of high burn-up fuels is essential to improve economy of the fast reactor fuel cycle. Increase of fuel burn-up is known to cause fuel-cladding chemical interaction (FCCI) and it mainly determines a lifetime of fuel pin. In order to extend a lifetime of fuel pin by mitigating FCCI, development of low oxygen-to-metal (O/M) MOX fuel has been carried out in plutonium fuel development center of JAEA. MOX fuel needs adjustment of the O/M ratio to less than 1.97 for high burn-up of 150 GWd/t. Therefore, O/M adjustment technology is one of the main subjects in development of a simplified MOX pellet fabrication method which has been advanced in the FaCT (Fast reactor Cycle Technology development) project. In previous work, changes in O/M ratio of MOX pellet during heat treatment were calculated from measurement results of oxygen potentials. On the basis of above calculation, heating tests were carried out to prepare low O/M ratio MOX pellets on a laboratory scale. The O/M ratios obtained in the heating tests were well consistent with calculation results. In the present study, a kilogram MOX scale furnace to adjust O/M ratio of MOX pellets for targeted value has been developed as next step.
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.
Permana, S.; Suzuki, Mitsutoshi; Suud, Z.*; Saito, Masaki*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12
Spent fuel (SF) of light water reactors (LWR) is loaded to increase breeding capability and intrinsic aspect of proliferation resistance. This present study intends to evaluate the effect of different loaded doping materials in the blanket region of FBR to the reactor performance and plutonium proliferation resistance level. Basic reactor operation is adjusted to reach 800 days operation by adopting 4 fuel batches systems of Japan Sodium Fast Reactor (JSFR) design. LWR plutonium compositions of Isotopic 
Pu and 
Pu show more sensitive to the decay time due to it shorter half-life which affect to the vector composition of plutonium as well as minor actinide (MA) as a function of decay time.
Kojima, Kensuke; Okumura, Keisuke; Asai, Shiho; Hanzawa, Yukiko; Okamoto, Tsutomu; Toshimitsu, Masaaki; Inagawa, Jun; Kimura, Takaumi; Kaneko, Satoru*; Suzuki, Kensuke*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
Accurate inventory estimation of long-lived fission products (LLFPs) in LWR spent fuels is important for the quality management and for long-term safety assessment of high-level radioactive vitrified wastes. In Japan, ORIGEN2 has been widely used to estimate the fuel compositions. However, equipped library data in the original ORIGEN2 are old and are not validated enough for LLFPs, such as Se, 
Tc, 
Sn and 
Cs, because available post irradiation examination (PIE) data are limited for these nuclides, which have difficulties in radiochemical analyses. For more accurate the estimation, new ORIGEN2 libraries are developed from the latest nuclear data library JENDL-4.0 for cross sections and fission yields, and from other libraries for half-lives, and so on. The new libraries are validated by PIE analyses of the sample fuels irradiated in Cooper, Calvert-Cliffs-1, H. B. Robinson-2, and Ohi-1. As a result, it was found that the new library gives good results for the estimation.
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).
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.
Suzuki, Mitsutoshi; Demuth, S.*
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
Proliferation risk assessment has been investigated to develop a performance-based approach in which the likelihood of diversion incidence, as well as the uncertainty of nuclear material accounting, is simultaneously considered to install intrinsic and extrinsic countermeasures in a conceptual design of a future reprocessing facility. A simulation and modeling approach has been applied to evaluate safeguards performance in a facility-level and diversion pathway analysis, which is demonstrated to detect more efficiently a small, protracted diversion that is usually investigated by trend analysis.
Arai, Yasuo; Serizawa, Hiroyuki; Nakajima, Kunihisa; Takano, Masahide; Sato, Isamu; Katsuyama, Kozo; Akie, Hiroshi; Suzuki, Motoe; Shirasu, Noriko; Haga, Yoshinori; et al.
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
High amount of He is generated in MA-bearing fuel during irradiation and storage periods compared with that in U or U-Pu fuel. Laboratory scale experiments, post irradiation examinations and modeling study were carried out in order to understand the He behavior in MA-bearing oxide fuel. Diffusion characteristics of He in single-crystal UO were investigated by the Knudsen effusion mass spectrometry. Effects of the He accumulation on lattice and bulk expansion of oxide pellets were examined by use of alpha-decay of 
Cm. Post irradiation examinations of 0.5%Am-MOX fuel irradiated at a fast test reactor JOYO were carried out, concentrating on the He behavior in the fuel pellets. A model describing the He behavior in MA-MOX fuel was constructed based on the principle processes, such as generation, diffusion, equilibrium and release to outer gaseous phase. By use of the model as a subroutine of a conventional fuel behavior analysis code, the He behavior in MA-MOX fuel for fast reactors was simulated.
Osaka, Masahiko; Konashi, Kenji*; Hayashi, Hirokazu; Li, D.*; Homma, Yoshiya*; Yamamura, Tomoo*; Sato, Isamu; Miwa, Shuhei; Sekimoto, Shun*; Kubota, Takumi*; et al.
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
Summer schools for future experts have successfully been completed under Japan Actinide Network (J-ACTINET) for the purpose of development of human resources who are expected to be engaged in every areas of actinide-research/engineering. The first summer school was held in Ibaraki-area in August 2009, followed by the second one in Kansai-area in August 2010. Two summer schools have focused on actual experiences of actinides in actinide-research fields for university students and young researchers/engineers as an introductory course of actinide-researches. Several quasi actinide-handling experiences at the actinide-research fields have attracted attentions of participants at the first school in Ibaraki-area. The actual experiments using actinides-containing solutions have been carried out at the second school in Kansai-area. Future summer schools will be held every year for the sustainable human resource development in various actinide-research fields.
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.
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.