Hirooka, Shun; Matsumoto, Taku; Kato, Masato; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*
Journal of Nuclear Materials, 542, p.152424_1 - 152424_9, 2020/12
The measurement of oxygen potential was conducted at 1,673, 1,773, and 1,873 K for (UPuAm)O and at 1,873 and 1,923 K for (UPuAmNp)O by using a thermo-gravimeter and an oxygen sensor. Am inclusion in terms of substituting the U significantly increased the oxygen potential. Similarly, the inclusion of Np as a substitute for U increased the oxygen potential; however, the effect was not as large as that with the Pu or Am addition at the same rate. The results were analyzed via defect chemistry and certain defect formations were suggested in the reducing region and the near-stoichiometric region by plotting the relationship between PO and the deviation from the stoichiometry. The equilibrium constants of the defect reactions were arranged to reproduce the experiment such that Am/Np contents were included in the entropy with coefficients fitting the experimental data.
Watanabe, Masashi; Kato, Masato; Sunaoshi, Takeo*
Journal of Nuclear Materials, 542, p.152472_1 - 152472_7, 2020/12
The oxygen self-diffusion coefficients in near stoichiometric (U,Pu)O at high temperatures were successfully measured by thermogravimetry combined with the oxygen isotope exchange method. The activation energy for oxygen diffusion in the stoichiometric composition of (U,Pu)O was evaluated from experimental data, and the value was determined to be 248 kJ/mol. In addition, the defect migration energies of (U,Pu)O were derived, and the oxygen self-diffusion coefficients were evaluated using these. As a result, good agreement was found between the experimental data and the oxygen self-diffusion coefficients calculated using the defect migration energies.
Segawa, Tomoomi; Kawaguchi, Koichi; Ishii, Katsunori; Suzuki, Masahiro; Fukasawa, Tomonori*; Fukui, Kunihiro*
Funtai Kogakkai-Shi, 57(9), p.485 - 494, 2020/09
In the spent fuel reprocessing process, a mixed solution of uranyl nitrate and plutonium nitrate is converted into mixed oxide powder by the microwave heating. To evaluate the applicability to the industrial-scale and acquire the characteristics data of the microwave heating denitration of various metal nitrate aqueous solutions based on the knowledge studied in the development of laboratory-scale basic experiments, the microwave heating characteristics and metal oxide powder properties were investigated using cerium nitrate, cobalt nitrate and copper nitrate aqueous solutions. The progress rate of the denitration reaction was depended on the position, and the denitration reaction proceeded faster at the periphery than at the center. The morphologies of the synthesized products were porous and hard dry solid with cerium nitrate aqueous solution, foamed dry solid with cobalt nitrate aqueous solution, and powdery particles with copper nitrate aqueous solution. The denitration ratio and average particle size of the synthesized products increased in the order of the cerium nitrate aqueous solution, the cobalt nitrate aqueous solution, and the copper nitrate aqueous solution. The numerical simulations revealed that the periphery of the bottom surface of the metal nitrate aqueous solution was heated by microwaves. This results consistent with the experimental results in which the denitration reaction started from the periphery of the metal nitrate aqueous solution.
Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Morimoto, Kyoichi; Kato, Masato
2018 GIF Symposium Proceedings (Internet), p.315 - 320, 2020/05
Recently, a research group studying at Plutonium Fuel Development Facility (PFDF) in Japan Atomic Energy Agency has systematically measured vast amounts of physical properties in the non-stoichiometric (U, Pu)O. Lattice parameter, elastic modulus, thermal expansion, oxygen potential, oxygen chemical diffusion coefficient and thermal conductivity were successfully measured as function of Pu content, O/M ratio and temperature, and the effects of Pu content and O/M ratio on their physical properties were evaluated. In this work, those experimental data are reviewed, and latest experimental data set on the non-stoichiometric (U, Pu)O are presented. The data set would be available in development of a fuel performance code.
Ikusawa, Yoshihisa; Hirooka, Shun; Uno, Masayoshi*
2018 GIF Symposium Proceedings (Internet), p.321 - 327, 2020/05
Research and development of Minor actinides (MAs) bearing MOX fuel for fast reactor has been proceeding from the viewpoint of reducing radioactive waste. In order to develop, MA bearing MOX, it is indispensable to clarify the influence of MA addition on irradiation behavior. The addition of Americium (Am) to MOX affects vapor pressure and thermal conductivity, which are important properties from the perspective of evaluating fuel temperature. This is because vapor pressure affects fuel restructuring, and thermal conductivity affects fuel temperature distribution. Focusing on these physical properties, this study evaluates the influence of Am on fuel temperature using irradiation behavior analysis code to contribute to the development of MA-bearing MOX fuel. An increase in Am content decreases the thermal conductivity and increases the oxygen potential of oxide fuel. Because vapor pressure increases with increasing Am content, pore migration is accelerated, and the central void diameter increases with increasing Am content. As a result, after formation of the central void, the influence of Am content on the fuel center temperature is mild. Alpha particles generated by radioactive decay of transuranium elements cause lattice defects in the oxide fuel pellets. It is well known that this phenomenon, which is called self-irradiation, affects thermal conductivity. Since americium is the typical alpha radioactive nucleus, to evaluate fuel temperature of Am-MOX is necessary to take account of the influence of self-irradiation damage on thermal conductivity. Self-irradiation decreases thermal conductivity, and as the Am content increases, the rate of decrease in thermal conductivity is accelerated. Because it recovers with temperature rise, the decrease in thermal conductivity due to self-irradiation damage has very little effect on fuel center temperature. These results suggest that Am-MOX fuel could be irradiated under the same conditions as conventional MOX fuel.
Segawa, Tomoomi; Yamamoto, Kazuya; Makino, Takayoshi; Iso, Hidetoshi; Kawaguchi, Koichi; Ishii, Katsunori; Sato, Hisato; Fukasawa, Tomonori*; Fukui, Kunihiro*
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.738 - 745, 2019/09
In the MOX fuel fabrication process, the dry grinding technology of mixed oxide pellets have been developed for the effective use of nuclear fuel materials. To develop a technology to control the particle size of dry recovered powder, the performance of the buhrstone mill and the collision plate type jet mill were studied using a simulated powder of particle size distribution about 500 m. We found that the particle size can be controlled at the range of about 250 m or less by both by adjusting the clearance between the grinding wheels of the buhrstone mill, and the clearance and elevation angle of the clarification zone of the the collision plate type jet mill. And furthermore, the collision plate type jet mill is considered to be suitable for particle size control because the operating parameters of the classifier can be finely adjusted.
Yamada, Yoshikazu; Shibanuma, Kimikazu
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.648 - 654, 2019/09
During a periodic inspection, multiple spot-like nuclear material contamination (maximum 21.7 Bq) was detected at the outer surface of a glove-box (GB) body used to install equipment for fabricating mixed oxide (MOX) fuel at the Japan Atomic Energy Agency. The inspection confirmed a total of 13 cracks passing through the thickness direction of the GB and a bleeding phenomenon was observed on the polyvinyl chloride (PVC) cables in the GB. These cracks were judged as stress corrosion cracking induced by the generation of chlorine gas by irradiation of PVC cables lying against the inner surface of the GB.
Segawa, Tomoomi; Kawaguchi, Koichi; Kato, Yoshiyuki; Ishii, Katsunori; Suzuki, Masahiro; Fujita, Shunya*; Kobayashi, Shohei*; Abe, Yutaka*; Kaneko, Akiko*; Yuasa, Tomohisa*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05
A solution of plutonium nitrate and uranyl nitrate is converted into a mixed oxide by microwave heating denitration method. In the present study, for improving the efficiency of microwave heating and achieving high-temperature uniformity to produce homogeneous UO powder, the microwave heating test of potassium chloride and uranyl nitrate solution, and numerical simulation analysis were conducted. The potassium chloride agar was adjusted to the dielectric loss, which is close to that of the uranyl nitrate solution and the optimum support table height was estimated to be 50 mm for denitration of the uranyl nitrate solution by microwave heating. The adiabator improved the efficiency of microwave heating denitration. Moreover, the powder yield was improved by using the adiabator owing to ease of scraping of the denitration product from the bottom of the denitration vessel.
Suzuki, Kiichi; Kato, Masato; Sunaoshi, Takeo*; Uno, Hiroki*; Carvajal-Nunez, U.*; Nelson, A. T.*; McClellan, K. J.*
Journal of the American Ceramic Society, 102(4), p.1994 - 2008, 2019/04
The fundamental properties of CeO were assessed using a range of experimental techniques. The oxygen potential of CeO was measured by the thermogravimetric technique, and a numerical fit for the oxygen potential of CeO is derived based on defect chemistry. Mechanical properties of CeO were obtained using sound velocity measurement, resonant ultrasound spectroscopy and nanoindentation. The obtained mechanical properties of CeO are then used to evaluate the Debye temperature and Gruneisen constant. The heat capacity and thermal conductivity of CeO were also calculated using the Debye temperature and the Gruneisen constant. Finally, the thermal conductivity was calculated based upon laser flash analysis measurements. This result demonstrates that the thermal conductivity has strong dependence upon material purity.
Ikusawa, Yoshihisa; Morimoto, Kyoichi; Kato, Masato; Saito, Kosuke; Uno, Masayoshi*
Nuclear Technology, 205(3), p.474 - 485, 2019/03
This study evaluated the effects of plutonium content and self-irradiation on the thermal conductivity of mixed-oxide (MOX) fuel. Samples of UO fuel and various MOX fuels were tested. The MOX fuels had a range of plutonium contents, and some samples were stored for 20 years. The thermal conductivity of these samples was determined from thermal diffusivity measurements taken via laser flash analysis. Although the thermal conductivity decreased with increasing plutonium content, this effect was slight. The effect of self-irradiation was investigated using the stored samples. The reduction in thermal conductivity caused by self-irradiation depended on the plutonium content, its isotopic composition, and storage time. The reduction in thermal conductivity over 20 years' storage can be predicted from the change of lattice parameter. In addition, the decrease in thermal conductivity caused by self-irradiation was recovered with heat treatment, and recovered almost completely at temperatures over 1200 K. From these evaluation results, we formulated an equation for thermal conductivity that is based on the classical phonon-transport model. This equation can predict the thermal conductivity of MOX fuel thermal conductivity by accounting for the influences of plutonium content and self-irradiation.
Funtai Kogakkai-Shi, 55(10), P. 547, 2018/10
no abstracts in English
Fujita, Shunya*; Abe, Yutaka*; Kaneko, Akiko*; Yuasa, Tomohisa*; Segawa, Tomoomi; Yamada, Yoshikazu; Kato, Yoshiyuki; Ishii, Katsunori
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Mixed uranium oxide and plutonium oxide powder is produced from uranyl nitrate and plutonium nitrate mixed solution by the microwave heating denitration method in the spent fuel reprocessing process. Since the microwave heating method is accompanied by a boiling phenomenon, it is necessary to fully grasp the operating conditions in order to avoid flashing and spilling in the mass production of denitrification technology for the future. In this research, it was confirmed that a potassium chloride aqueous solution as a simulant of uranyl nitrate aqueous solution with high dielectric loss cause loss of microwave at the solution surface as the dielectric loss increased with the increase of KCl concentration by experimental and electromagnetic field analysis, and revealed that the change in the heating condition affects the generation of flushing.
Hirooka, Shun; Kato, Masato; Watanabe, Masashi
Transactions of the American Nuclear Society, 118, p.1624 - 1626, 2018/06
This study suggested the time development of oxygen-to-metal ratio (O/M) redistribution model with oxygen-related properties in MOX. Irradiation simulation including the suggested O/M redistribution and pore migration with vaporization-condensation model which bares density redistribution was demonstrated. The simulation results showed that O/M redistribution proceeded at lower temperature than density redistribution, which indicated that oxygen diffusion got influential at lower temperature than vaporization-condensation of MOX. Another find was that O/M redistribution was very slow at the surface because temperature kept low. However, near the surface (inside from the surface) where the temperature exceeded 1000 K, O/M redistribution was rather recognizable with oxygen flown from inner region to the near-surface. The results will be evaluated by comparison with post-irradiation examination data.
Hirooka, Shun; Kato, Masato
Journal of Nuclear Science and Technology, 55(3), p.356 - 362, 2018/03
The sound speeds of longitudinal and transverse waves in the uranium-plutonium mixed oxide (MOX) pellets were measured as functions of porosity, oxygen-to-metal ratio (O/M) and plutonium content. The effect of each parameter was well fitted by a linear function and the equations were obtained to calculate the sound speeds. Mechanical properties were evaluated with the sound speeds and the result of Young's modulus showed that porosity was the most important factor to decrease Young's modulus. Temperature dependence on Young's modulus was also evaluated with previously reported thermal expansion. Decrease of Young's modules with increasing temperature was in good agreement with available literature.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
In a recycle system for minor actinides (MAs) currently studied to reduce the degree of hazard and the amount of high-level radioactive wastes, MAs will be recycled by reprocessing and irradiating as mixed oxide (MOX) with plutonium (Pu) and uranium (U) in a fast reactor. MA content is expected to be 5 wt.% in the future recycle system, and MAs might affect irradiation behavior of MA-MOX fuels. The main influences of MA-containing would be increase of fuel temperature and cladding stress, and the important behaviors would be fuel restructuring, redistribution, helium (He) generation and cladding corrosion. The MA-containing influences were evaluated with CEPTAR.V2, including fuel properties and analysis models to evaluate the MA-MOX fuel irradiation behavior, by using the results of highly americium (Am) containing MOX irradiation experiment, B8-HAM, performed in Joyo. The irradiation behavior of Am-MOX fuels could be precisely analyzed and revealed the influences of Am-containing.
Fujita, Shunya*; Abe, Yutaka*; Kaneko, Akiko*; Chonan, Fuminori*; Yuasa, Tomohisa*; Yamaki, Tatsunori*; Segawa, Tomoomi; Yamada, Yoshikazu
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
From the observation results, in the process of flushing, the behaviors leading to flushing were classified divided into three types. First type is that first generation bubble from heating leads to flushing. Second type is that nucleate boiling continues during heating and stop, finally single bubble generates and leads to flushing. Third type is defined that gradual evaporation occurs without bubbles. It was revealed that the total quantities of heat released by flushing are approximately equal when assuming the flushing mechanism, it can be triggered that a large amount of micro bubbles are instantaneously generated and grew.
Hirooka, Shun; Kato, Masato
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 6 Pages, 2017/06
Young's modulus of MOX pellets was evaluated by measuring the sound velocities of longitudinal and transverse waves in the pellets as functions of porosity, oxygen-to-metal ratio (O/M) and plutonium content. The results showed that porosity was the most important factor that 20% of the porosity decreased Young's modulus by neatly 100 GPa while O/M and plutonium content could change the Young's modulus by ~20 GPa. From the measured sound velocities, temperature dependence on Young's modulus and specific heat capacity were calculated on the Debye model by leveraging the thermal expansion data. The temperature dependence that Young's modulus decreases with increasing temperature is in good agreement with literature data. The specific heat capacity also agrees with that of calculated value by Kopp's method, taken the Schottky term and the excited term into account.
Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.
Watanabe, Masashi; Sunaoshi, Takeo*; Kato, Masato
Defect and Diffusion Forum, 375, p.84 - 90, 2017/05
The oxygen chemical diffusion coefficient in (U, Pu)O was determined by thermo-gravimetry as functions of the Pu content, oxygen-to-metal ratio and temperature. The surface reaction was considered in the diffusion coefficient determination. The activation energy for the chemical diffusion coefficient was 60 kJ/mol and 65 kJ/mol, respectively, in (UPu)O and (UPu)O.
Kato, Masato; Nakamura, Hiroki; Watanabe, Masashi; Matsumoto, Taku; Machida, Masahiko
Defect and Diffusion Forum, 375, p.57 - 70, 2017/05
The basic properties of PuO were reviewed, and the equilibrium defects in PuO were evaluated from the experimental data of the oxygen potential and electrical conductivity as well as the Ab-initio calculation results. Consistency among various properties was confirmed, and the mechanistic models for thermal property representations were derived.