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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.
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.
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.
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.
Hayashi, Hirokazu; Sato, Takumi; Shibata, Hiroki; Iwai, Takashi; Nishihara, Kenji; Arai, Yasuo
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
R&D on the transmutation of long-lived minor actinides (MA) by the accelerator-driven system (ADS) using nitride fuels is underway at Japan Atomic Energy Agency. In regard to reprocessing technology, pyrochemical process has several advantages in case of treating spent fuel with large decay heat and fast neutron emission, and recovering highly enriched N-15. In the pyrochemical reprocessing, plutonium (Pu) and MA are dissolved in LiCl-KCl eutectic melts and selectively recovered into liquid cadmium (Cd) cathode by molten salt electrorefining. The recovered Pu and MA are converted to nitrides by the nitridation-distillation combined method, in which the Cd alloys containing Pu and MA are heated in nitrogen gas stream. The authors have investigated its elemental technologies such as electrorefining and renitridation. On the other hand, development of the process flow diagram with the material balance sheet of the pyrochemical reprocessing of spent nitride fuel for ADS is in progress. This paper summarized recent progress of the study which aims to prove the technological applicability of pyrochemical process to the nitride fuel cycle for transmutation of MA.
Miwa, Shuhei; Osaka, Masahiko; Usuki, Toshiyuki; Sato, Isamu; Tanaka, Kosuke; Hirosawa, Takashi; Yoshimochi, Hiroshi; Onose, Shoji
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
A new fast reactor (FR) cycle concept was previously proposed that incorporates MgO-based inert matrix fuels (IMFs) containing minor actinides harmonious with the existing FR cycle technologies. A basic study of MgO-based IMFs was made regarding their fabrication, characterization and reprocessing in terms of applicability to existing FR cycle technology. It was concluded from these basic investigations of MgO-based IMFs that the existing FR cycle technologies can be applied to those for MgO-based IMFs, and the basic technologies of MgO-based IMFs containing minor actinides harmonious with the existing FR cycle technologies were established.
-dodecane and HNO
systemSasaki, Yuji; Kitatsuji, Yoshihiro; Tsubata, Yasuhiro; Sugo, Yumi; Shirasu, Noriko; Morita, Yasuji
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
Mutual separation of Am, Cm and lanthanides (Ln) is important to develop the partitioning process of high-level radioactive waste, although this method is difficult to establish due to their very similar chemical behavior, same oxidation state (III) and similar ionic radii. Relative high separation values for these metals can be seen when the multidentate ligands, e.g., diethylene-triamine-N,N,N',N'',N''-pentaacetic acid (DTPA), and softN-donor extractants are employed. We examine to use the combination of two-multidentate ligands having N atoms in their structures in behalf of the high separation of Am, Cm and Ln. Various diamide-type ligands can be synthesized from the initial materials of carboxylic acid and amine, and the aimed materials can be tailored to have high solubility in either HNO orn-dodecane, in case short or long alkyl chains are attached in amidic Natoms. Taking the simultaneous use of these materials into consideration, hydrophilic and lipophilic diamide compounds with different complexing ability for trivalent Ln and actinide (An) are used into both of the aqueous and the organic phases in order to enlarge their separation factors. This technique is applicable in the salt-free system and diamide can be gasificated by combustion, which reduces the secondary-radioactive waste. Two diamide compounds, DGA (diglycolamide) and DOODA (dioxaoctanediamide) as the representative multidentate ligands, have the different features for Ln-complexation, namely DGA has higher D values for middle-Ln than those of light-Ln, and DOODA shows the opposite trend to DGA. TDdDGA (N,N,N',N'-tetradodecyl-diglycolamide) and DOODA(C2) (N,N,N',N'-tetraethyl-dioxaoctanediamide) have high solubility in n-dodecane and HNO, respectively, and the condition of 0.1 M TDdDGA in n-dodecane and 0.5 MDOODA(C2) in 3 M HNO gives the separation factors of La/Gd: 1128 and Am/Cm:3.26.
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.
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.
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
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.
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.
Okano, Masanori; Jitsukata, Shu*; Kuno, Takehiko; Yamada, Keiji
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12
Nakamura, Kinya*; Ogata, Takanari*; Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Kato, Tetsuya*; Arai, Yasuo; Koyama, Tadafumi*; Itagaki, Wataru; Soga, Tomonori; et al.
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
CRIEPI and JAEA have fabricated sodium-bonded metallic fuel elements for the first time in Japan as a collaborative research, for use in the irradiation test at the experimental fast test reactor Joyo. The irradiation test aims to assess the irradiation behavior of the fuel and the internal wastage of the stainless-steel cladding by rare-earth fission products at a maximum cladding temperature above 873 K. U-20 wt% Pu-10 wt% Zr alloy fuel slugs of 200 mm length were fabricated in an injection-casting furnace using U metal, U-Pu alloy and Zr metal. Two types of fuel slug were fabricated, i.e., 5.05 mm and 4.95 mm in diameter, and loaded into a ferritic-martensitic stainless-steel cladding tubes, respectively. After top-end-plug welding to the cladding tube, each fuel element was subjected to sodium bonding to fill the annular gap between the fuel slug and the cladding with melted sodium. The fabrication results indicated that the characteristics of the fuel elements were within the required specifications.