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Takahatake, Yoko; Watanabe, So; Irisawa, Keita; Shiwaku, Hideaki; Watanabe, Masayuki
Journal of Nuclear Materials, 556, p.153170_1 - 153170_7, 2021/12
Times Cited Count：1 Percentile：15.7(Materials Science, Multidisciplinary)Suzuki, Eriko; Nakajima, Kunihisa; Miwa, Shuhei; Osaka, Masahiko; Hashimoto, Naoyuki*; Isobe, Shigehito*
no journal, ,
In order to improve the cesium (Cs) chemisorption model onto stainless steel (SUS), the chemical forms and distributions of Cs compounds formed inside of the SUS oxide layer at 873-1273 K were evaluated based on the microscopic observation. It was revealed that compositions of Cs compounds varies alonbg depth from the oxide layer surface on SUS, which was tested in 1073-1273 K. In addition, an amorphous phase containing Cs was observed inside of the oxide layer on the SUS tested in 873-973 K. These results suggest that the chemical reactions occurred inside of the SUS oxide layer are different from that in the nearest region of the SUS oxide layer surface.
Ozawa, Takayuki; Hirooka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*
no journal, ,
To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with INL and have installed pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation experiment in Joyo was evaluated by the 2-dimensional analyses. Four MA-bearing MOX pins with different O/M ratio and pellet/cladding gap size were irradiated in Joyo B14 experiment. Remarkable restructuring of stoichiometric MA-bearing MOX fuels was observed in PIE, and could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void assumes to move toward wide-gap side when the pellet eccentricity taking place, but 2-dimentional analyses on pellet transverse section revealed that the central void formation observed in PIE would be inconsistent with a direction of the pellet eccentricity.
Ishii, Katsunori; Segawa, Tomoomi; Kawaguchi, Koichi; Nishina, Masahiro; Makino, Takayoshi; Natori, Yuri*
no journal, ,
JAEA is developing simplified a plutonium and uranium mixed oxide (MOX) pellet fabrication process. In this process, agitation granulator improves flowability of MOX powder with water as binder. This granulation method has issues, including low production capacity due to criticality control for wet nuclear material. A new simple additive-free dry granulation method was proposed recently to produce tritium breeding Li
O spheres for nuclear fusion reactor. In this research, results of experiments to investigate the applicability of the new granulation method to MOX powder is reported.
Segawa, Tomoomi; Kawaguchi, Koichi; Ishii, Katsunori; Yamamoto, Kazuya; Makino, Takayoshi; Iso, Hidetoshi; Fukasawa, Tomonori*; Fukui, Kunihiro*
no journal, ,
Japan Atomic Energy Agency has been used out of specification mixed oxide (MOX) pellets as a dry recovered powder for the effective use of nuclear fuel material in the MOX fuel fabrication process. The densities of the sintered MOX pellets can be controlled to about 85 %T.D. without adding pore former by adjusting the amount and the particle size of the dry recovered powder into the raw powder. It is required to adjust the particle size of the dry recovered powder to under 250 
m, the influence of the operating parameters of the collision plate-type jet mill on the characteristics of pulverization and the influence of pulverized powders on sintering properties were evaluated. The clearance was narrowed, the pulverized powders were confirmed to be adjusted for the particle diameter of under 250 m, and the pellet prepared from the pulverized powder with density of about 85.0 %T.D. was obtained.
Furukawa, Tomohiro; Takai, Toshihide; Watanabe, Shigeki*; Ishioka, Noriko*
no journal, ,
Authors have started the development of liquid bismuth target system for continuous production of therapeutic radionuclide astatine-211 as an application of liquid metal technology cultivated in above research and development, in recent years. Fe-Cr-Al alloy is one of the candidates of the target window in this system. In order to clarify the corrosion of the target window in pure liquid bismuth, the exposure test at elevated temperature was done for Fe-Cr-Al alloys and their references. The test was conducted under argon gas flow, i.e. saturated dissolved oxygen condition in theory. The temperature and exposure time were 500
C and 500 hours, respectively. Results of metallurgical examination after the exposure, it was concluded that the corrosion behavior was basically equivalent to that in lead-bismuth eutectic, which is the candidate coolant of heavy metal cooled fast reactors and accelerator driven systems.
Ikusawa, Yoshihisa; Ozawa, Takayuki; Kato, Masato
no journal, ,
MOX is one of the most typical nuclear fuel. MOX fuel is divided into two categories. One is low Pu contained MOX fuel, which is used in light water reactors, with Pu content of less than 15 wt%. The other is high plutonium contained MOX fuel, which is used in fast reactors, with Pu content of more than 20 wt%. The physical properties of MOX fuel are depended on Pu content. In addition, the irradiation behavior of MOX fuel is significantly different between LWRs and FRs due to the difference in these operating environments. Therefore, various studies of MOX fuel have been conducted separately for LWR and FR. In our facility, MOX physical properties, such as thermal conductivity, melting point, and specific heat, have been measured using various MOX fuel samples. Based on these measurements, a mechanistic MOX fuel property model has been developed. This model can evaluate the physical properties of MOX fuel for any specification. Irradiation behavior analysis codes developed based on this property model might have the potential to be a general-purpose code that can evaluate the behavior of any MOX fuel. In this study, as a first step of development of a general-purpose irradiation behavior analysis code, the fuel temperature analysis function of the irradiation analysis code "DIRAD" was confirmed by using the data of high Pu contained MOX fuel irradiated in JOYO and low Pu contained MOX fuel irradiated in Halden. As the result, it was confirmed that the fuel temperature of low to high Pu contained MOX fuel could be evaluated by using DIRAD code adopting the property model.
Yokoyama, Keisuke; Watanabe, Masashi; Kato, Masato; Tokoro, Daishiro*
no journal, ,
In current nuclear fuel cycle systems, to reduce the amount of high-level radioactive waste, minor actinides (MAs) bearing MOX fuel is one option for burning MAs using fast reactor. However, the effects of Am content in fuel on thermal conductivity are unclear because there are no experimental data on thermal conductivity of high Am bearing MOX fuel. In this work, the thermal conductivity of high Am-bearing MOX fuel samples was measured. In this study, MOX fuel samples containing 10 at% Am were prepared. The oxygen to metal ratio (O/M ratio) of sintered pellet was adjusted to 2.00. Thermal diffusivity was measured from R.T. to 1473 K by the laser flash method. Thermal conductivity was calculated from thermal diffusivity, heat capacity and the density of fuel samples. The measured thermal conductivity values decreased with the increase of Am content. Those for 10 at% Am bearing MOX fuel agreed well with the classical phonon transport model, and the effects of bearing 10 at% Am on MOX fuel samples were in good agreement with those predicted from previous experimental study results.
Nishi, Tsuyoshi*; Ota, Hiromichi*; Kokubo, Hiroki*; Sato, Rika*; Yamano, Hidemasa
no journal, ,
In this study, the viscosities of the molten SS (SUS316L), 2.5mass%B
C-SS, 5.0mass%BC-SS, 7.0mass%BC-SS and 10mass%BC-SS alloys were measured by the oscillating crucible method. The viscosity measurements of the molten SS, 2.5mass%BC-SS and 5mass%BC-SS, 7mass%BC-SS and 10mass%BC-SS alloys were performed in the temperature range from 1613 to 1793 K, from 1713 to 1793 K, and from 1793 to 1823 K, respectively. In these results, the viscosity increased with BC components in the BC components range from 0 to 7.0mass%. The equation of viscosity of molten BC-SS alloys was estimated by using the experimental data of the molten 2.5mass%BC-SS and 5.0mass%BC-SS, and 7.0mass%BC-SS in the temperature range from 1713 to 1793 K. The equation of viscosity of molten SS+BC alloys was determined as follows. y=Ax+B, A=0.0302, B=-9.881
10
T+2.546, where x is the BC component and T is the temperature. The uncertainty of the equation is 2.5%.
Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa
no journal, ,
Thermophysical properties of molten mixture of 316L stainless steel (SS) and control-rod material (BC) are necessary for the development of computer simulation codes that describe core degradation mechanisms during severe accidents in nuclear power plants involving sodium-cooled fast reactors. To satisfy this demand, the present authors first measured the liquidus temperatures of the SS-BC mixtures up to 10 mass% of BC by using differential scanning calorimetry (DSC). Based on these data, the thermophysical properties of the molten SS-BC mixtures were measured by using a noncontact high-temperature thermophysical property measurement system, which consists of an electromagnetic levitator, a superconducting magnet, a laser-heating system, a high-speed charge-coupled-device camera, a data-logging system, and a gas-control system. The density, surface tension, normal spectral emissivity, specific heat capacity, and thermal conductivity of molten mixtures were measured over a wide temperature range with high accuracy in a noncontact way. This paper provides the liquidus temperatures and thermophysical properties recently updated in the project.
