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Someya, Takayuki*; Chitose, Hiromasa*; Watanabe, Satoshi*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
In this study, CFD analysis has been conducted for the assessment of spent fuel integrity in large LOCA event and the maximum temperature of spent fuel assemblies has been evaluated. Then, it has been compared with the result of the simple assessment method. As a case study, additional CFD analysis has been conducted, where water level in SFP decreases to the Bottom of Active Fuel (BAF) due to boil-off. Since this scenario might be more severe than large LOCA scenario, the number of spent fuel assemblies, their decay heat and loading pattern to maintain spent fuel integrity are investigated.
Someya, Yoji; Tobita, Kenji; Yanagihara, Satoshi*; Kondo, Masatoshi*; Uto, Hiroyasu; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Sakamoto, Yoshiteru
Fusion Engineering and Design, 89(9-10), p.2033 - 2037, 2014/10
Times Cited Count：9 Percentile：53.95(Nuclear Science & Technology)In the replacement period of a fusion power reactor, the assembly of blanket or divertor modules need to be removed from the reactor in order to minimize remote maintenance in the vacuum vessel and to attain a reasonable plant availability. In the hot cell, the modules will be removed from the backplate of the assembly. Here, note that the active cooling must be done by a way that does not cause contamination of the hot cell environment due to dispersion of tritium and tungsten dust. In this sense, the cooling scenario is adopted that the existing pipe of cooling water in the assembly is connected to a different cooling water system in the hot cell. In this scenario, the temperature of the assembly is maintained about 40-100
C. On the other hand, the structural material (RAFM) of the blanket and divertor is not recycled due to its high contact dose rate. It should be crushed into small pieces to reduce volume of the waste and required storage space. Here, the decay heat must be removed by natural convection to keep the temperature below 65C for preventing water evaporation from the mortar. The RAFM is kept in the interim storage during 12 years until the required temperature conditions for mortar are ensured and then is disposed of.
Hirose, Takanori; Someya, Yoji; Tanigawa, Hisashi; Suzuki, Satoshi
Yosetsu Gakkai-Shi, 83(1), p.70 - 77, 2014/01
no abstracts in English
Tobita, Kenji; Uto, Hiroyasu; Kakudate, Satoshi; Takase, Haruhiko; Asakura, Nobuyuki; Someya, Yoji; Liu, C.
Fusion Engineering and Design, 86(9-11), p.2730 - 2734, 2011/10
Times Cited Count：13 Percentile：67.37(Nuclear Science & Technology)For high availability of DEMO operation, sector horizontal transport hot cell maintenance scheme was studied. Transport of sector with 730 tons is carried out using a wheeled platform. The driving force of pulling the sector into a cask is ball screws. The fulcrum of the ball screws is the cryostat wall so that a large pulling force is expected with no-counter balance. The cask containing the sector is delivered by air casters from the cryostat to the hot cell. For the maintenance scheme, new concepts such as transfer of the tilting forces of toroidal coils using ropes and shafts and supports for the tilting force using reinforced concrete floor or cryostat wall were proposed. Based on the maintenance concept, the period required for replacement of all sectors is estimated to be 35.5-67.5 days, satisfying the design target (shorter than 3 months).
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Sato, Satoshi; Seki, Yohji; Takase, Haruhiko
Fusion Engineering and Design, 86(9-11), p.2378 - 2381, 2011/10
Times Cited Count：12 Percentile：61.50(Nuclear Science & Technology)For DEMO reactor blanket design, a two-dimensional (2-D) nuclear-thermal-coupled analysis code, DOHEAT, has been developed. In DOHEAT, the neutron flux is calculated by a 2-D transport code, DOT3.5, with the nuclear data library, FUSION-40, and the nuclear heating rate and the local TBR profile of blanket are calculated using the 2-D neutronics calculation code, APPLE-3. Use of the code has showed outstanding usefulness in the blanket design where detailed evaluation of neutron flux, nuclear heating rate, tritium breeding ratio (TBR) and the temperature of materials is required for various blanket concepts and trial-and-error-basis iteration is sometimes necessary. DOHEAT can replace the actual blanket structure by a more realistic model including cooling tubes, multipliers and breeders. A validation calculation indicates that DOHEAT provides reasonable results on the temperature profile.
Ishihara, Ryo*; Fujiwara, Kunio*; Harayama, Takato*; Okamura, Yusuke*; Uchiyama, Shoichiro*; Sugiyama, Mai*; Someya, Takaaki*; Amakai, Wataru*; Umino, Satoshi*; Ono, Tsubasa*; et al.
Journal of Nuclear Science and Technology, 48(10), p.1281 - 1284, 2011/10
Times Cited Count：44 Percentile：93.82(Nuclear Science & Technology)Someya, Satoshi*; Uchida, Mitsunori*; Uchibori, Akihiro; Ohshima, Hiroyuki; Li, Y.*; Okamoto, Koji*
Nihon Kikai Gakkai Rombunshu, B, 75(759), p.2173 - 2181, 2009/11
When the pressurized water or water vapor leaks from a failed heat transfer tube in a steam generator of sodium cooled fast reactors, the high-velocity and high-temperature jet with sodium-water chemical reaction may cause wastage of the adjacent tubes. The evaluation of the wastage rate requires numerical analysis of the reaction jet. To construct a numerical method for the reaction jet, experiment on the underexpanded gas jet into the liquid pool need to be done. In this study visualization and measurement for the underexpanded nitrogen gas jet injected into the water was carried out. The horizontal penetration length of the gas jet and the expansion angle were obtained from the averaged images of a high-speed camera. The experimental results showed that the penetration length and the expansion angle increased approximately linearly with increasing stagnation pressure. The entrainment velocity and the velocity of the entrained water droplets were obtained by PIV.
Uchida, Mitsunori*; Someya, Satoshi*; Okamoto, Koji*; Ohshima, Hiroyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/09
When a heat exchanger in fast breeder reactor cracks, a sodium-water reaction occurs. Highly pressurized water or steam escapes into the surrounding liquid sodium. The release of steam into the liquid sodium media is a two-phase flow with an underexpansion. Several studies have examined only the underexpansion of the gas-gas phase. However, there are few reports on the underexpansion of the gas-liquid phase. In this study quantitative measurement was carried out for the purpose of revealing the flow with the underexpanded gas jet injected into water. The gas jet distance and the expansion angle were then obtained from the averaged images of a high-speed camera. The gas jet distance and the expansion angle increased approximately linearly with increasing pressure. The entrainment velocity and the velocity of entrained water droplets into the gas jet were obtained by PIV.
Kureta, Masatoshi; Kumada, Hiroaki; Kume, Etsuo; Someya, Satoshi*; Okamoto, Koji*
Proceedings of 3rd International Workshop on Process Tomography (IWPT-3) (CD-ROM), 8 Pages, 2009/04
Dynamic neutron tomography velocimetry has been developed in order to obtain the 3-D velocity distribution and flow profile data of liquid metal flow in a heated rod bundle for development of an advanced nuclear reactor. In this paper, data processing methods for the 3-D velocimetry is focused on. The data processing is started from the reading images recorded by the three high-speed video cameras, and is finished to the visualization of velocity of the tracers and the profiles. Basic experiments were carried out using the research reactor JRR-4 and the dynamic neutron tomography system. As the results, it was confirmed that the 3-D velocity distribution and flow profile could be visualized by the new data processing methods.
Kureta, Masatoshi; Kumada, Hiroaki*; Kume, Etsuo; Someya, Satoshi*; Okamoto, Koji*
Journal of Physics; Conference Series, 147, p.012087_1 - 012087_14, 2009/03
The aim of this development is to visualize and measure the velocity distribution in liquid metal flow using the neutron beam with the high-speed imaging technique, computer tomography (CT) technique and particle tracking velocimetry (PTV). Final research purpose is to obtain the velocity distribution and flow profile data of liquid metal flow in a heated rod bundle for development of the FBR core. In this paper, visualization and measurement method using the JRR-4, spring model PTV method for this technique and results of the fundamental PTV experiment were reported. The fundamental experiment was conducted. As the result, cadmium tracers buried in the aluminum column with the speed of 1.5 revolving per second could be visualized as the 3D movie under 125Hz and 250Hz sampling conditions, the profile of the tracer could be traced, and fundamental velocity distribution measurement method could be conformed.
Kureta, Masatoshi; Kumada, Hiroaki; Kume, Etsuo; Someya, Satoshi*; Okamoto, Koji*
Proceedings of 6th International Symposium on Measurement Techniques for Multiphase Flows (ISMTMF 2008) (USB Flash Drive), 14 Pages, 2008/12
The aim of this development is to visualize and measure the velocity distribution in liquid metal flow using the neutron beam with the high-speed imaging technique, computer tomography (CT) technique and particle tracking velocimetry (PTV). Final research purpose is to obtain the velocity distribution and flow profile data of liquid metal flow in a heated rod bundle for development of the FBR core. In this paper, visualization and measurement method using the JRR-4, spring model PTV method for this technique and results of the fundamental PTV experiment were reported. The fundamental experiment was conducted. As the result, cadmium tracers buried in the aluminum column with the speed of 1.5 revolving per second could be visualized as the 3D movie under 125 Hz and 250 Hz sampling conditions, the profile of the tracer could be traced, and fundamental velocity distribution measurement method could be conformed.
Ochi, Daisuke*; Someya, Satoshi*; Ohshima, Hiroyuki; Okamoto, Koji*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
It is important to advance safety and efficiency of the FBR. A numerical prediction method of high accuracy and its validation with a simulated thermal flow field are indispensable. A wire-wrapped rod bundle system, which briefly simulated the fuel rods system in FBR, was built up in the experiments. The wire-wrapped rods were made from Mexflon-material, of which refractive index was exactly same with that of water. The particle image velocimetry was applied to measure the velocity field in the narrow gap between wire-wrapped rods, under variable flow rates with or without heating rods. The aim of this study was to contribute to the certification of results of numerical simulation for the safety design of the FBR.
Someya, Satoshi*; Ochi, Daisuke*; Ohshima, Hiroyuki; Okamoto, Koji*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
Temperature sensitive particles incorporating phosphor molecules were synthesized. These particles, suitable for particle image velocimetry, were used to measure the velocity and temperature distributions in water flowing through a wire-wrapped rod bundle system, which simulated the fuel rod system in a fast breeder reactor. The particles were illuminated by a pulse laser at 20 Hz. A high speed camera was used to record 30 particle images at intervals of 25
50 micro seconds (2040 kHz) for each excitation laser pulse. From each series of images the velocity and temperature fields were calculated. This measurement technique should contribute to the experimental validation of numerical simulations for the safe design of fast breeder reactors.
Uchida, Mitsunori*; Someya, Satoshi*; Okamoto, Koji*; Ohshima, Hiroyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
When a heat exchanger in a fast breeder reactor cracks, a sodium-water reaction occurs. When a tube cracks, highly pressurized water or steam escapes into the surrounding liquid sodium. The release of steam into the liquid sodium media is a two-phase flow with an underexpansion. There have been few reports on the underexpansion of the gas-liquid phase. In this paper, qualitative measurement of the two-phase flow was carried out for the purpose of revealing the flow with the underexpanded gas jet injected into water. The gas jet range and the gas jet width were then obtained from averaged images of a high-speed camera. PIV was also carried out by observing scattering light from the gas bubbles. The gas jet range and the gas jet width increased approximately linearly with increasing pressure. The results of PIV showed that the bubble velocity increased increasing pressure.
Someya, Satoshi*; Okamoto, Koji*; Ohshima, Hiroyuki; Yoshida, Satoshi*; Li, Y.*
no journal, ,
In this study, simultaneous measurement of temperature and velocity was conducted successfully by using temperature sensitive particles.
Okamoto, Koji*; Someya, Satoshi*; Kureta, Masatoshi; Kumada, Hiroaki; Kume, Etsuo
no journal, ,
no abstracts in English
Nakamura, Makoto; Tobita, Kenji; Tanigawa, Hisashi; Someya, Yoji; Masui, Akihiro; Watanabe, Kazuhito; Konishi, Satoshi*; Torikai, Yuji*
no journal, ,
Tritium is major radioactive material in a fusion reactor. Evaluation of the dose due to the tritium is essential to understand environmental consequences of incidental or accidental conditions postulated in the fusion reactor. A purpose of this study is to identify issues to apply UFOTRI, a code of tritium dose analysis being used for the ITER safety assessment, the Japanese environmental conditions. Extensive scans of UFOTRI calculation runs were performed in various meteorological and release conditions. The scans show that the contribution of the secondary tritium release is more significant in the cases of lower release height, lesser stable atmosphere or more distant conditions. The analysis, thus, suggests that it is important to take into account the contribution of the secondarily released tritium in evaluating the early dose to the public due to the tritium release.
Kureta, Masatoshi; Kumada, Hiroaki; Kume, Etsuo; Someya, Satoshi*; Okamoto, Koji*
no journal, ,
Development of the Dynamic Neutron Computer Tomography (DNCT) technique is conducted as the new 4D thermal-hydraulic measurement technique in order to provide the detailed database for thermal-hydraulic evaluation methods on liquid metal cooled FBR core. We studied the data analysis method to be realized a velocity measurement by the DNCT, and developed the velocity and locus measurement techniques by expanding the spring model particle trace velocimetry. In this presentation, velocity measurement technique installed in the DNCT analysis system and test results taken at the JRR-4 are reported.
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Kakudate, Satoshi; Tanigawa, Hisashi; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke
no journal, ,
Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. In this study, we categorize various schemes in term of [1] the maintenance port position for transporting blanket segments, [2] blanket segmentation, and [3] divertor segmentation. The design study clarifies some assessment factors on DEMO remote maintenance scheme, for example, (1) minimization of the size and magnetic stored energy of TF coil and PF coils, (2) divertor maintenance. This presentation describes engineering design of each maintenance schemes and evaluation results of comparison among maintenance schemes.
Watanabe, Kazuhito; Nakamura, Makoto; Someya, Yoji; Masui, Akihiro; Katayama, Kazunari*; Hayashi, Takumi; Yanagihara, Satoshi*; Konishi, Satoshi*; Yokomine, Takehiko*; Torikai, Yuji*; et al.
no journal, ,
In the DEMO design, the blanket primary cooling system involves high temperature pressurized water (~300C). This means the temperature of blanket structural material is higher than that of ITER. This increases tritium permeation ratio from the fusion plasma and blanket breeder to the primary cooling water. Therefore, we need to consider installation of a water detritiation system. In this study, we estimate the demand of water detritiation system from the view point of the amount of tritium permeated to primary cooling water that assumed conservatively. We also organize the issues for management of tritiated water from the other point of view based on the characteristic of the fusion DEMO reactor. The result shows that the existing facilities can be adopted to the DEMO if we can control the tritium ratio of primary cooling water as same as that of CANDU reactor.
