Matsuba, Kenichi; Kato, Shinya; Kamiyama, Kenji; Akayev, A. S.*; Baklanov, V. V.*
Proceedings of 28th International Conference on Nuclear Engineering; Nuclear Energy the Future Zero Carbon Power (ICONE 28) (Internet), 4 Pages, 2021/08
In order to obtain experimental knowledge on fragmentation and cooling behavior of molten core material discharged into regions where the depth and volume of sodium are limited, a series of out-of-pile experiments using molten alumina as a simulant for molten core material was conducted. It was found that following mechanisms might be involved in the fragmentation and cooling behavior in a shallow sodium pool: (1) FCI which occurs at location of impingement of the molten jet on the bottom plate promotes fragmentation. (2) If there is a sufficient amount of sodium as a heat sink outside the region, heat exchange by sodium flow in and out due to vapor expansion and condensation suppresses the sodium temperature rise. (3) This temperature suppression contributes to effective cooling of molten core material. In the future study, in order to confirm the mechanisms which was clarified in this study, analytical evaluation of the experimental result will be carried out using a simulation tool.
Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
Nakamichi, Shinya; Hiroka, Shun; Kato, Masato; Sunaoshi, Takeo*; Nelson, A. T.*; McClellan, K. J.*
Journal of Nuclear Materials, 535, p.152188_1 - 152188_8, 2020/07
Oxygen-to-metal ratio (O/M) of uranium and plutonium mixed oxide depends on its oxygen partial pressure. To attain the desirable microstructure and O/M ratio of sintered pellets, it is important to investigate the relation between the sintering behavior and the atmosphere of sintering process. In this study, sintering behavior of (PuU)O and (PuU)O in controlled po atmosphere were investigated. It was found activation energy of (PuU)O was higher than that of (PuU)O. On the other hand, it was observed grain growth during sintering was suppressed in hypo-stoichiometric composition.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05
In order to ensure In-Vessel Retention (IVR) of molten-core in Core Disruptive Accident (CDA), we are investigating the possibility of the molten-core discharge through the control rod guide tube (CRGT) to prevent energetics due to exceeding the prompt criticality. Internal structures of the CRGT, such as a sodium-flow regulator when the CRGT is connected to the high-pressure plenum, may disturb the discharge of molten-core from the core region. Based on above background, an experimental program to clarify characteristics of molten-core discharge through the CRGT has been commenced as one of subjects under a joint study with National Nuclear Center of the Republic of Kazakhstan (NNC-RK) named EAGLE-3 project. An experiment using molten-alumina as fuel simulant and sodium was conducted at the out-of-pile test facility owned by NNC-RK to investigate sodium cooling effect around the sodium flow regulator on its destruction. The experimental result represented that void development at the initiation of molten-alumina discharge eliminated liquid-phase sodium from the discharge path and this also eliminated sodium cooling effect around the sodium flow regulator. As a result, early destruction of the sodium flow regulator and massive discharge of molten alumina occurred in turn.
Nakamichi, Shinya; Hiroka, Shun; Sunaoshi, Takeo*; Kato, Masato; Nelson, A.*; McClellan, K.*
Transactions of the American Nuclear Society, 113(1), p.617 - 618, 2015/10
Cerium dioxide has been used as a surrogate material for plutonium dioxide. Dorr et al reported the use of hyper-stoichiometric conditions causes the start of shrinkage of (U,Ce)O at low temperature compared with the sintering in reducing atmosphere. However, the precise stoichiometry of the samples investigated was not controlled or otherwise monitored, preventing any quantitative conclusions regarding the similarities or differences between (U,Ce)O and (U,Pu)O. The motivation for the present work is therefore to compare the sintering behavior of MOX and the (U,Ce)O MOX surrogates under controlled atmospheres to assess the role of oxygen defects on densification in both systems.
Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro
JAEA-Technology 2014-043, 36 Pages, 2015/02
The core management and operation code system aims to perform core management task efficiently by systematic management of data, analyses and edits, which are needed in the reactor core management and operation. The system consists of the five calculation modules: the reactor constant generation module, the neutronic-thermal calculation module, the radiation analysis module, the core structural integrity estimation module, and the core operation analysis module. In these modules, the neutronic-thermal calculation module is based on the dedicated three-dimensional diffusion and burn-up code HIZER. HIZER can execute core calculations easily for specific design specification and operation patterns of Monju, enabling efficient and accurate evaluation of the Monju core characteristics. This report describes its calculation method and validation results.
Arifi, E.*; Ishimatsu, Koichi*; Iizasa, Shinya*; Namihira, Takao*; Sakamoto, Hiroyuki*; Tachi, Yukio; Kato, Hiroyasu*; Shigeishi, Mitsuhiro*
Construction and Building Materials, 67(Part B), p.192 - 196, 2014/09
The Fukushima Dai-ichi Nuclear Plant accident has resulted in a large amount of radioactively contaminated concrete. The possible application of the pulsed power discharge to reduce the amount of contaminated concrete as radioactive waste was investigated. The contaminated concrete was decontaminated by separating contaminated matrix from uncontaminated coarse aggregate under pulsed power discharge process. In this study, a stable Cs isotope was used to simulate radioactively contaminated concrete. As a result, while the volume of reclaimed aggregate from contaminated concrete could be reproduced was up to 60%, nevertheless Cs detected in the reclaimed aggregate was only approximately 3%. Thus most of the Cs were dissolved in water during the discharge process. It is expected that the pulsed power could reduce the contaminated concrete waste by reusing aggregate. Further investigations are requested to test the applicability of this method under the realistic conditions close to the actual waste.
Hirano, Fumio; Sato, Seichi*; Kozaki, Tamotsu*; Inagaki, Yaohiro*; Iwasaki, Tomohiko*; Oe, Toshiaki*; Kato, Kazuyuki*; Kitayama, Kazumi*; Nagasaki, Shinya*; Niibori, Yuichi*
Journal of Nuclear Science and Technology, 49(3), p.310 - 319, 2012/03
The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the fuel histories including the burn-up of UO spent fuels, the cooling period before reprocessing, the storage period of fresh MOX fuels. The heat generation rates of hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 GWd-MOX needs to be limited to a value of 0.7 to 1.6, which is significantly lower than the value of 4.0 for 45 GWd-UO.
Kato, Masato; Nakamichi, Shinya; Takeuchi, Kentaro; Sunaoshi, Takeo*
CALPHAD; Computer Coupling of Phase Diagrams and Thermochemistry, 35(4), p.623 - 626, 2011/12
Uranium and plutonium mixed oxide (MOX) has been used as fuels of fast reactors. The MOX having fluorite structure is an oxygen nonstoichiometric compound which is stable in hyper- and hypo-stoichiometric composition range. The stoichiometry of MOX significantly affects their properties. So, it is essential to know the relation between stoichiometry and oxygen potential to develop MOX fuels. In this work, the oxygen potentials of (UPu)O were measured at high temperatures of 1773, and 1873K. The measurements were carried out by gas equilibrium method using thermo-gravimetry. Th The oxygen partial pressure was adjusted by controlling the ratio of H/HO in the flowing gas atmosphere, and the oxygen potential was determined. The oxygen potentials were determined as functions of O/M ratio, and temperature. The data at stoichiometric composition were estimated to be -311kJ/mol and -299kJ/mol at 1773K, and 1873K based on point defect model.
Nakamichi, Shinya; Kato, Masato; Tamura, Tetsuya*
CALPHAD; Computer Coupling of Phase Diagrams and Thermochemistry, 35(4), p.648 - 651, 2011/12
Japan Atomic Energy Agency has developed homogeneous MOX fuels containing minor actinides (MAs) such as Am and Np elements. In this work, particular attention is paid to the influence of small MA addition on oxygen potential of MOX fuels. Oxygen potentials of (AmPuU)O and (AmNpPuU)O in the near stoichiometric region were measured at 1473, 1573 and 1623K by thermogravimetry. (AmPuU)O and (AmNpPuU)O were slightly higher than those of (PuU)O. From this work, it was found that oxygen potential of MOX is slightly increased by small Am addition and influence of Np addition on oxygen potential is not significant.
Sakanaka, Shogo*; Akemoto, Mitsuo*; Aoto, Tomohiro*; Arakawa, Dai*; Asaoka, Seiji*; Enomoto, Atsushi*; Fukuda, Shigeki*; Furukawa, Kazuro*; Furuya, Takaaki*; Haga, Kaiichi*; et al.
Proceedings of 1st International Particle Accelerator Conference (IPAC '10) (Internet), p.2338 - 2340, 2010/05
Future synchrotron light source using a 5-GeV energy recovery linac (ERL) is under proposal by our Japanese collaboration team, and we are conducting R&D efforts for that. We are developing high-brightness DC photocathode guns, two types of cryomodules for both injector and main superconducting (SC) linacs, and 1.3 GHz high CW-power RF sources. We are also constructing the Compact ERL (cERL) for demonstrating the recirculation of low-emittance, high-current beams using above-mentioned critical technologies.
Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki
Proceedings of 10th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation (CD-ROM), p.201 - 209, 2010/00
Japan Atomic Energy Agency has developed homogeneous MOX fuel containing minor actinide (MA) elements such as Np and Am. To measure physical properties of the fuel is essential for its development, because their data are needed to evaluate irradiation behavior. In this report, the physical properties, melting temperature, thermal conductivity, lattice parameter, oxygen potential and phase separation behavior, were reviewed, and effect of MA content was discussed.
Nakamichi, Shinya; Kato, Masato; Sunaoshi, Takeo*; Uchida, Teppei; Morimoto, Kyoichi; Kashimura, Motoaki; Kihara, Yoshiyuki
Journal of Nuclear Materials, 389(1), p.191 - 196, 2009/05
Japan Atomic Energy Agency researchers have developed mixed oxide (MOX) fuels containing minor actinides (MA). These fuels were irradiated for ten minutes in the FBR Joyo in some short-term irradiation tests. The Si-condensed phases were observed at the center of the pellets in the post irradiation examination. Si impurities came to be mixed into the raw materials in the ball milling process, because Si rubber was used as the lining of the milling pot. Content of Si in the pellets was within the specification of the fuel. It is important to investigate the Si state in MOX at high temperatures like the reactor operating temperature of the fuel to evaluate irradiation behavior. In the present work, MOX specimens with mixed SiO impurity were prepared. The ratio of MOX to SiO was controlled at a mol fraction of 3 to 1. The specimens were first heated at 1973K in atmospheres of three different oxygen partial pressures to adjust the O/M ratio. Then these specimens were sealed in a tungsten capsule, and heated at 2273K or 2673K. Compounds consisting of Pu and Si were observed at grain boundaries of the MOX matrix in specimens after heat treatment. These compounds were not observed in grain interior and MOX matrix was not affected significantly by Si impurity. These compounds tended to form in specimens with low O/M ratio and in specimens heated at higher temperatures.
Kato, Masato; Morimoto, Kyoichi; Nakamichi, Shinya; Sugata, Hiromasa*; Konashi, Kenji*; Kashimura, Motoaki; Abe, Tomoyuki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(4), p.420 - 428, 2008/12
The melting temperatures of MOX for fast reactor fuel were investigated as functions of Pu content, Am content and oxygen-to-metal (O/M) ratio using thermal arrest technique. Rhenium inner was used for the measurement to prevent the reaction between the sample and capsule materials. The solidus temperatures decreased with increasing Pu and Am content and increased with decreasing O/M ratio. It is considered that the maximum temperature in U-Pu-O system varies in hypostoichiometric composition region. The melting temperatures were evaluated by ideal solid solution model in UO-PuO-AmO-PuO system, and the model was derived for calculating solidus and liquidus temperature. The derived model reproduced the experimental data with 25 K.
Inoshita, Shinya*; Suzuki, Shogo*; Okada, Yukiko*; Kato, Masahiko*; Hirai, Shoji*; Kimura, Atsushi; Hatsukawa, Yuichi; Toh, Yosuke; Koizumi, Mitsuo; Oshima, Masumi
Tetsu To Hagane, 94(9), p.345 - 350, 2008/09
"Tatara" is Japanese original steel making method. Steel made by "Tatara" is famous as low alloy and suited to "Kaji" process. By authors' study, it turned out that we could estimate the source region of raw material of Tatara by As and Sb concentration ratio in Tatara sample. But the concentration of these element in Tatara sample is very low (ppm or sub-ppm order), therefore, quantitative analysis is very difficult. In this study, we adopted Neutron Activation Analysis combined with Multiple -ray detection (NAAMG) to analyze As and Sb in "Tatara" sample (iron lump and sand iron, slag). NAAMG is high sensitive and non-destructive analysis method which combined NAA (Neutron Activation Analysis) and multiple -ray detectors. Each "Tatara" sample (iron lump, sand iron, slag) were irradiated for 1-2 h (for As measurement), 8-17 h (for Sb measurement) in JRR-3M HR irradiation field (thermal neutron flux was about 9.010 n/m s). And cooling time was 4-5 days (As), and 19-36 days (Sb). Coincidence -rays were measured by -ray detector array, GEMINI-II. Counting time was 1-8 hours (As), and 2-41 hours (Sb). Quantification was made by comparison method. As a result of measurement, the concentration of As and Sb in all "Tatara" samples were determined by NAAMG and these were sub-ppm order. Lower Limit of Determination (LLD) of As was 0.1 ppm order and Sb is 0.01ppm order. From the above-mentioned point, the effectiveness of NAAMG to analyze trace element in "Tatara" sample was confirmed.
Kato, Chiaki; Motooka, Takafumi; Numata, Masami; Endo, Shinya; Yamamoto, Masahiro
Structural Materials for Innovative Nuclear Systems (SMINS), p.439 - 447, 2008/07
Corrosion of a nuclear fuel reprocessing plant is also important problem in either current or an advanced nuclear fuel reprocessing system. In this process, nitric acid solution is used to dissolve spent nuclear fuels and solvent extraction method is used to separate U, Pu and actinoid elements. It will be much severer corrosive environment. In this paper, an effect of neptunium ions on corrosion of ultra low carbon Type 304 stainless steel was investigated. The corrosion tests were conducted in 9 kmol/m nitric acid solution adding neptunium ions. The results show that neptunium ions promote inter-grainier corrosion of SUS304ULC in nitric acid solution and corrosion rate in heat-transfer condition is larger than that in immersed condition. It is estimated that the oxidise potential of nitric ions increases under heat-transfer condition more than immersion condition in boiling solution. Furthermore, the effect of -ray irradiation used Co is examined. -ray irradiation decreases corrosion rates and the reason is discussed.
Sakanaka, Shogo*; Ago, Tomonori*; Enomoto, Atsushi*; Fukuda, Shigeki*; Furukawa, Kazuro*; Furuya, Takaaki*; Haga, Kaiichi*; Harada, Kentaro*; Hiramatsu, Shigenori*; Honda, Toru*; et al.
Proceedings of 11th European Particle Accelerator Conference (EPAC '08) (CD-ROM), p.205 - 207, 2008/06
Future synchrotron light sources based on the energy-recovery linacs (ERLs) are expected to be capable of producing super-brilliant and/or ultra-short pulses of synchrotron radiation. Our Japanese collaboration team is making efforts for realizing an ERL-based hard X-ray source. We report recent progress in our R&D efforts.
Kato, Masato; Nakamichi, Shinya; Takano, Tatsuo
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.916 - 920, 2007/09
no abstracts in English
Nakamichi, Shinya; Kato, Masato; Morimoto, Kyoichi; Sugata, Hiromasa*; Kashimura, Motoaki; Abe, Tomoyuki
Transactions of the American Nuclear Society, 96(1), p.191 - 192, 2007/06
JAEA has developed plutonium and uranium mixed oxide (MOX) containing 20-32%Pu content as a fuel of the fast breeder reactor. During irradiation, large temperature gradient in radial direction of a fuel pellet causes redistribution of Pu and U, and the Pu content increases to about 43% at the pellet center. The maximum temperature of the fuel pellet during irradiation is limited within the design criterion to prevent fuel melting. So, it is important to evaluate melting points of MOX containing 43%Pu. In this work, it is confirmed that the MOX with 43%Pu content is not melted by heat treatment just below the melting point which was determined by thermal arrest technique using Re inner capsule. The MOX specimen with 43%Pu content was heated at 2978K for 40s using Re inner capsule. Optical micrograph and XRD results show the specimen was heated at the temperature less than solidus temperature. So it was confirmed that (PuAmU)O was solid phase at 2978K20K.
Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Sugata, Hiromasa*; et al.
JAEA-Technology 2006-049, 32 Pages, 2006/10
Japan Atomic Energy Agency has developed a fast breeder reactor(FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio(O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Nuemann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content.