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Tsujimoto, Kazufumi; Arai, Yasuo; Minato, Kazuo
Nihon Genshiryoku Gakkai-Shi ATOMO
, 59(11), p.644 - 648, 2017/11
no abstracts in English
Arai, Yasuo
Bunri Henkan Gijutsu Soron, p.134 - 146, 2016/08
no abstracts in English
Arai, Yasuo
Genshiryoku, Ryoshi, Kakuyugo Jiten, 3, p.61 - 67, 2014/12
no abstracts in English
Hayashi, Hirokazu; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki; Iwai, Takashi; Arai, Yasuo
Science China; Chemistry, 57(11), p.1427 - 1431, 2014/11
Times Cited Count：5 Percentile：5.84(Chemistry, Multidisciplinary)Nitride fuels have several advantages, such as high thermal conductivity and high metal density like metallic fuels, and high melting point and isotropic crystal structure like oxide fuels. Since the late 1990s, the partitioning and transmutation of minor actinides (MA) has been studied to decrease the long term radio-toxicity of high level waste and mitigate the burden on the final disposal. Japan Atomic Energy Agency (JAEA) has been proposing dedicated transmutation cycle using the Accelerator-Driven System (ADS) with the nitride fuels containing MA. We have been developing the nitride fuel cycle including pyrochemical process. Our focus is on electrolysis of nitride fuels and refabrication of nitride fuel from the recovered actinides because other processes are similar to the technology for the metal fuel treatment and have been studied elsewhere. In this paper, we summarized our activity on developments of the pyrochemical treatment of the spent nitride fuels.
Shibata, Hiroki; Hayashi, Hirokazu; Akabori, Mitsuo; Arai, Yasuo; Kurata, Masaki
Journal of Physics and Chemistry of Solids, 75(8), p.972 - 976, 2014/08
Times Cited Count：18 Percentile：59.26(Chemistry, Multidisciplinary)Gibbs free energies of formation of six Ce-Cd intermetallic compounds, CeCd, CeCd
, CeCd
, CeCd
, CeCd
and CeCd
, were evaluated systematically using electrochemical techniques in the temperature range from 673 to 923 K in the LiCl-KCl-CeCl-CdCl molten salt bath. The linear dependence of the Gibbs free energies of formation on temperature yields to the enthalpies and entropies of formation of these intermetallic compounds. By extrapolating the molar Gibbs free energy of Ce-Cd intermetallic compounds to the Cd distillation temperature, it was clear that the molar Gibbs free energy of Ce in Ce-Cd intermetallic compounds decreases gradually from CeCd to CeCd and attains to the minimum value at CeCd. This suggests on the Cd distillation from the U-Pu-Ce-Cd alloy that the dissolution of U or Pu into CeCd should be mostly taken into consideration.
Nishi, Tsuyoshi; Arai, Yasuo; Takano, Masahide; Kurata, Masaki
JAEA-Data/Code 2014-001, 45 Pages, 2014/03
JAEA-Data-Code-2014-001.pdf:3.57MBJAEA-Data-Code-2014-001(errata).pdf:0.2MB
The purpose of this study is to prepare a property database of nitride fuel needed for the fuel design of accelerator-driven system (ADS) for transmutation of minor actinide (MA). Nitride fuel of ADS is characterized by high content of Pu and MA as principal components, and addition of a diluent material such as ZrN. Experimental data or evaluated values from the raw data on properties Pu and MA nitrides, and nitride solid solutions containing ZrN are collected and summarized, which cover the properties needed for the fuel design of ADS. They are expressed as an equation as much as possible for corresponding to a variety conditions. Error evaluation is also made as much as possible. Since property data on transuranium (TRU) nitrides are often lacking, those on UN and (U,Pu)N are substitutionally shown in such cases in order to facilitate the fuel design with a tolerable accuracy by complementing the database.
Nishi, Tsuyoshi; Takano, Masahide; Arai, Yasuo; Kurata, Masaki
Dai-34-Kai Nihon Netsu Bussei Shimpojiumu Koen Rombunshu, p.199 - 201, 2013/11
By installing the laser flash apparatus and the drop calorimeter in the glove box, the thermal diffusivity and the heat capacity measurements of nitride containing MA elements of long-lived radioactive nuclides were enabled. The sample holder and the platinum container were designed to measure the thermal diffusivity and the heat capacity of very small quantity of MA nitride samples. The thermal conductivities of MA nitride increased with temperature, unlike that of conventional oxide-type nuclear fuels. In addition, the thermal conductivities of MA nitride decreased with increasing Am contents. The thermal conductivity of ZrN-based MA nitride, which is proposed as a candidate material for the ADS fuel, was fitted to equations as functions of the temperature and ZrN concentration. The predicted values agreed well with the experimental ones, indicating that the thermal conductivity of nitride fuel for ADS can be predicted for a practical design.
Pu
Am
Cm
)O
solid solutionsNishi, Tsuyoshi; Takano, Masahide; Akabori, Mitsuo; Arai, Yasuo
Journal of Nuclear Materials, 440(1-3), p.534 - 538, 2013/09
Times Cited Count：2 Percentile：18.63(Materials Science, Multidisciplinary)To clarify the dependence of thermal conductivity on storage time of curium containing oxide, the authors prepared the sintered sample of (NpPuAmCm)O (x = 0.02, 0.04) solid solutions and evaluated the thermal conductivity. The thermal conductivities of (NpPuAmCm)O exponentially decreased with increasing storage duration. This result suggested that the degradation of the thermal conductivities was attributed to the accumulation of lattice defects by self-irradiation.
Cm
)ONishi, Tsuyoshi; Takano, Masahide; Akabori, Mitsuo; Arai, Yasuo
Journal of Nuclear Materials, 433(1-3), p.531 - 533, 2013/02
Times Cited Count：6 Percentile：44.02(Materials Science, Multidisciplinary)To clarify the storage duration dependence of the thermal conductivity of MA containing oxide fuel, the thermal diffusivity of (PuCm)O was measured at 473, 523 and 573 K by a laser flash method using the sample stored for 48, 264, 504, and 960 h. The heat capacity was measured by a drop calorimetry to derive the thermal conductivity. It was confirmed that the degradation of the thermal conductivity was attributed to the accumulation of lattice defects caused by self-irradiation, because the storage duration dependence of the thermal conductivity could be approximated by the equation used for self-irradiation lattice expansion model.
Akie, Hiroshi; Sato, Isamu; Suzuki, Motoe; Serizawa, Hiroyuki; Arai, Yasuo
Journal of Nuclear Science and Technology, 50(1), p.107 - 121, 2013/01
Times Cited Count：2 Percentile：18.63(Nuclear Science & Technology)A simple formula is developed for the evaluation of the helium production amount in the fast reactor fuel. For the subroutine use in the existing fuel behavior analysis code, the formula is designed putting emphasis on simplicity and quickness rather than accuracy. The accuracy of the formula is confirmed by comparing with the detailed calculation with SWAT code, and also with the post irradiation examination (PIE) results of the fuel pin irradiated at the experimental fast reactor JOYO. As a result, the formula is found to evaluate the helium amount with the difference of less than about 10% from the detailed calculation and from the PIE results. Based on these results, the formula is installed in the fuel behavior analysis code for the simulation of helium behavior in fast reactor fuels.
Haga, Yoshinori; Homma, Yoshiya*; Aoki, Dai*; Nakajima, Kunihisa; Arai, Yasuo; Matsuda, Tatsuma; Ikeda, Shugo*; Sakai, Hironori; Yamamoto, Etsuji; Nakamura, Akio; et al.
Journal of the Physical Society of Japan, 81(Suppl.B), p.SB007_1 - SB007_4, 2012/12
Times Cited Count：0 Percentile：0(Physics, Multidisciplinary)Arai, Yasuo
Comprehensive Nuclear Materials, 3, p.41 - 54, 2012/03
Development of nitride fuel was extensively reviewed in this paper, which will be published in the Comprehensive Nuclear Materials. Characteristics of nitride fuel and history of the R&D were described in the introductive chapter. In the chapter of fabrication, nitride fuel preparation by carbothermic reduction and its sintering behavior were centered on the chapter. In the chapter of irradiation behavior, results of the irradiation tests performed so far were reviewed and the in-pile behavior of nitride fuel was described. Furthermore, in the chapter of reprocessing, hydrochemical and pyrochemical reprocessing technologies of nitride fuel were briefly explained. Outlook of nitride fuel was given in the last chapter. On the other hand, physical, chemical and thermodynamic properties of nitride fuel will be reviewed in another part of the Comprehensive Nuclear Materials.
Arai, Yasuo
Procedia Chemistry, 7, p.425 - 430, 2012/00
Times Cited Count：1 Percentile：56.78(Chemistry, Analytical)The fundamental research on actinide materials has been carried out in order to contribute to the development of future nuclear fuel cycle and actinide science database. Among actinide materials, the R&D has been focused on Pu and minor actinide (MA; Np, Am, Cm) bearing compounds. The chemical forms of actinide compounds concerned include oxides, nitrides, chlorides and alloys, which are prepared, characterized and subjected to property measurements. In this paper those results on Pu and MA bearing oxides obtained in recent several years are summarized. In addition, the possible challenges of actinide materials research to the subjects of post severe accident of Fukushima Dai-ichi Nuclear Power Station are briefly discussed.
Sato, Takumi; Nishihara, Kenji; Hayashi, Hirokazu; Kurata, Masaki*; Arai, Yasuo
Proceedings of 11th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation (Internet), 9 Pages, 2012/00
Nitride fuel cycle for transmutation of long-lived minor actinides (MAs) has been developed in Japan Atomic Energy Agency (JAEA) under the double-strata nuclear fuel cycle concept. This study aims at developing the process flow diagram with the material balance sheet of the pyrochemical reprocessing of spent nitride fuel for ADS to evaluate the technological feasibility of the fuel cycle. Three process flow diagrams were proposed in this work: (1) the currently-proposed process with the molten salt electrorefining of spent nitride fuel, (2) the process with the molten salt electrorefining of the actinide-cadmium alloy after the chemical dissolution of spent nitride fuel, and, (3) the process combining the chemical dissolution of spent nitride fuel and the multi-stage counter current extraction. Moreover, the material balance was evaluated for the process (1) by use of the calculated and experimental data of the nitride fuel for ADS.
Nakamura, Kinya*; Kato, Tetsuya*; Ogata, Takanari*; Nakajima, Kunihisa; Iwai, Takashi; Arai, Yasuo
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
The first irradiation campaign of U-Pu-Zr metal fuel in Japan is planned in the experimental fast reactor JOYO. In the fabrication of U-Pu-Zr fuel, two methods were adopted for preparing U-Pu alloy from the oxide; one is the electrochemical reduction and the other is the electrorefining followed by reductive extraction. Injection casting for U-Pu-Zr slug was carried out after adding U and Zr metals to meet the target specifications of the irradiated fuel. Several conditions of Na-bonding process were determined from the results of tests using simulated metal fuel pins. Based on these results, six U-Pu-Zr fuel pins for the irradiation tests are now being fabricated.
Nakajima, Kunihisa; Serizawa, Hiroyuki; Shirasu, Noriko; Haga, Yoshinori; Arai, Yasuo
Journal of Nuclear Materials, 419(1-3), p.272 - 280, 2011/12
Times Cited Count：24 Percentile：85.64(Materials Science, Multidisciplinary)The solubility and diffusion coefficient of helium in the single-crystal UO samples were determined by a Knudsen-effusion mass-spectrometric method. The measured helium solubilities were found to lie within the scatter of the available data, but to be much lower than those for the polycrystalline samples. The diffusion analysis was conducted based on a hypothetical equivalent sphere model and the simple Fick's law. The helium diffusion coefficient was determined by using the pre-exponential factor and activation energy as the fitting parameters for the measured and calculated fractional releases of helium. The optimized diffusion coefficients were in good agreement with those obtained by a nuclear reaction method reported in the past. It was also found that the pre-exponential factors of the determined diffusion coefficients were much lower than those analyzed in terms of a simple interstitial diffusion mechanism.
Kikuchi, Hironobu; Nakamura, Kinya*; Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo; Ogata, Takanari*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.323 - 331, 2011/12
A high-purity Ar gas atmosphere glovebox accommodating injection casting and sodium-bonding apparatuses was newly installed in Plutonium Fuel Research Facility (PFRF) of Oarai Research and Development Center, Japan Atomic Energy Agency. Past experiences in PFRF led to the establishment of technological basis of fabrication of U-Pu-Zr alloy fuel pin for the first time in Japan. After the injection casting of U-Pu-Zr alloy, the metallic fuel pins are fabricated by welding upper- and lower end plugs with cladding tube of ferritic-martensitic steel. Subsequent to the sodium bonding for filling the annular gap region between the U-Pu-Zr alloy and cladding tube with the melted sodium, the fuel pins are subjected to the inspection for irradiation tests. This paper summarizes the equipment of the apparatuses and the technological basis for fabrication of U-Pu-Zr alloy fuel pins for the coming irradiation test in the experimental fast test reactor JOYO.
Nakamura, Kinya*; Ogata, Takanari*; Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Kato, Tetsuya*; Arai, Yasuo; Uozumi, Koichi*; Hijikata, Takatoshi*; Koyama, Tadafumi*; et al.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.245 - 256, 2011/12
Sodium-bonded metallic fuel elements were fabricated for the first time in Japan for the irradiation test in the experimental fast test reactor JOYO. U-20Pu-10Zr fuel slugs of 200 mm in length and approximately 5 mm in diameter were fabricated in a small-scale injection casting furnace. Each fuel slug was loaded into the ferritic martenstic stainless steel (PNC-FMS) cladding tube with the sodium thermal bond, thermal insulator and reflector in a helium gas atmosphere glove box. After top-end plug welding to the cladding tube and heat treatment of the welding area, each fuel element was subjected to the sodium bonding process. After the inspection such as element length, gas plenum length and helium-leak tightness, six metallic fuel elements are transported to the JOYO site for the coming irradiation test.
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.
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.
