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.
Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 11 Pages, 2019/05
The cooling of the residual core materials after the fuel discharge from the core in the accident of SFRs can significantly affect the distribution fraction of the core materials, which is an important factor for the in-vessel retention (IVR). For the evaluation of the cooling of the residual core materials which is called ``in-place cooling'', behavior in a SFR core was analyzed preliminary by SIMMER-III. Based on the analysis result, method of phenomena identification and ranking table (PIRT) was applied. Fundamental experiment focusing on three thermal-hydraulic phenomena those were extracted by PIRT was considered in order to investigate them and utilize it for validation of the SIMMER-III. To achieve continuous oscillation of sodium level which can occur in the phase of in-place cooling of SFRs, analytical survey was conducted by SIMMER-III. As a result of that, the effects of experimental conditions on the oscillation amplitude and the duration time were clarified quantitatively, which are necessary to determine the specific experimental conditions.
Kato, Masatsugu*; Funakoshi, Kanji*; Liu, X.*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11
Kurisaki, Tatsuya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11
In the evaluation of the in-place cooling which is for the residual core materials in the severe accident of sodium-cooled fast reactors, pressure loss of two-phase flow in debris bed is one of the important factors. Although Lipinski model is already proposed for the pressure loss evaluation, the accuracy would decrease when the porosity is not homogeneous. Thus, experiment to measure the pressure loss in a packed bed of non-homogeneous porosity distribution was conducted, and the Lipinski model was modified dividing the cross section to evaluate the pressure loss in it. As a result, it was confirmed that agreement of the experimental values with the values by modified Lipinski model was better than that with the original Lipinski model.
Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 4 Pages, 2018/11
In Core Disruptive Accidents of Sodium-cooled Fast Reactors, molten core material would be discharged through control rod guide tubes into the inlet coolant plenums beneath the rector cores. The inlet coolant plenums have quite limited heights and sodium inventories. Therefore, in the inlet plenums, molten core material with a jet-like shape would impinge on the bottom of the plenum before it breaks up into fragments. In this study, to clarify fragmentation behavior in a shallow sodium pool whose height and volume are so limited that jet impingement on the bottom is expected, an out-of-pile experiment discharging molten alumina into a sodium pool was conducted. Although a small amount of alumina agglomeration was found on the bottom plate (steel disk) installed in the sodium pool, most of the molten alumina was fragmented into debris particles. Results obtained in the present experiment suggest that molten core material is fragmented and quenched even in a shallow sodium pool.
Morita, Koji*; Ogawa, Ryusei*; Tokioka, Hiromi*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10
The EAGLE in-pile ID1 test has been performed by Japan Atomic Energy Agency to demonstrate early fuel discharge from a fuel subassembly with an inner duct structure, which is named FAIDUS. It was deduced that early duct wall failure observed in the test was initiated by high heat flux from the molten pool of fuel and steel mixture. The posttest analyses suggest that molten pool-to-duct wall heat transfer might be enhanced effectively by the molten steel with large thermal conductivity in the pool without the presence of fuel crust on the duct wall. In this study, mechanisms of heat transfer from the molten pool to the duct wall was analyzed using a fully Lagrangian approach based on the finite volume particle method for multi-component, multi-phase flows. A series of pin disruption, molten pool formation and duct wall failure behaviors was simulated to investigate mixing and separation behavior of molten steel and fuel in the pool, and their effect on molten pool-to-duct wall heat transfer. The present 2D particle-based simulations demonstrated that large thermal load beyond 10 MW/m on the duct wall was caused by effective heat transfer due to direct contact of liquid fuel with nuclear heat to the duct wall.
Phan, L. H. S.*; Ohara, Yohei*; Kawata, Ryo*; Liu, X.*; Liu, W.*; Morita, Koji*; Guo, L.*; Kamiyama, Kenji; Tagami, Hirotaka
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10
Self-leveling behavior of core fuel debris beds is one of the key phenomena for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The SIMMER code has been developed for CDA analysis of SFRs, and the code has been successfully applied to numerical simulations for key thermal-hydraulic phenomena involved in CDAs as well as reactor safety assessment. However, in SIMMER's fluid-dynamics model, it is always difficult to represent the strong interactions between solid particles as well as the discrete particle characteristics. To solve this problem, a new method has been developed by combining the multi-fluid model of the SIMMER code with the discrete element method (DEM) for the solid phase to reasonably simulate the particle behaviors as well as the fluid-particle interactions in multi-phase flows. In this study, in order to validate the multi-fluid model of the SIMMER code coupled with DEM, numerical simulations were performed on a series of self-leveling experiments using a gas injection method in cylindrical particle beds. The effects of friction coefficient on the simulation results were investigated by sensitivity analysis. Though more extensive validations are needed, the reasonable agreement between simulation results and corresponding experimental data preliminarily demonstrates the potential ability of the present method in simulating the self-leveling behaviors of debris bed. It is expected that the SIMMER code coupled with DEM is a prospective computational tool for analysis of safety issues related to solid particle debris bed in SFRs.
Wu, P.*; Zhang, B.*; Peng, K. L.*; Hagiwara, Masayuki*; Ishikawa, Yoshihisa*; Kofu, Maiko; Lee, S. H.*; Kumigashira, Hiroshi*; Hu, C. S.*; Qi, Z. M.*; et al.
Physical Review B, 98(9), p.094305_1 - 094305_7, 2018/09
Using angle-resolved photoemission spectroscopy and inelastic neutron scattering, we have studied how electronic structures and lattice dynamics evolve with temperature in Na-doped SnSe.
Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Matsuba, Kenichi; Kamiyama, Kenji
Nuclear Engineering and Design, 334, p.90 - 95, 2018/08
Two-phase flow through porous media must be well understood to develop a severe accident analysis code not only for light water reactor but also sodium-cooled fast reactor. When a core disruptive accident occurs in sodium-cooled fast reactor, the fuel inside the core become melted and interacts with the coolant. As a result, gas-liquid two-phase flow will be formed in the debris bed, which may have porous nature depending on the cooling process. In such condition, the local porosity and its distribution are very important to characterize two-phase flow field in the porous media. In this study, X-ray radiography was applied to measure the local porosity in the packed bed of spheres. The radial profiles were estimated from the chordal profiles measured by the X-ray method and compared with the previous porosity model. In addition, the void fraction radial profiles were also obtained in air-water two-phase flow.
Aoyagi, Mitsuhiro; Kamiyama, Kenji; Tobita, Yoshiharu
Journal of Nuclear Science and Technology, 55(5), p.530 - 538, 2018/05
Sheikh, Md. A. R.*; Son, E.*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Matsuba, Kenichi; Kamiyama, Kenji; Suzuki, Toru*
Journal of Nuclear Science and Technology, 55(6), p.623 - 633, 2018/01
During core-disruptive accidents in sodium-cooled fast reactors, the sedimentation behavior of fragmented debris is crucial for in-vessel retention. The height of the beds formed may influence both the cooling of the bed and the neutronic characteristics. To develop an experimental database of bed formation behavior, a series of experiments using simulant materials, namely, AlO, ZrO, and stainless steel, were performed under gravity-driven discharge of solid particles from a nozzle into a quiescent cylindrical water pool. The bed height was measured for particles of different size, density, and sphericity, and an injection nozzle with varying diameter, injection velocity, and injection height. From these experiments, an empirical correlation was established to predict the bed height for both homogeneous and mixed particles for the different properties. This correlation reproduces reasonably well the experimental trend in bed height.
Nakajima, Kenji; Kawakita, Yukinobu; Ito, Shinichi*; Abe, Jun*; Aizawa, Kazuya; Aoki, Hiroyuki; Endo, Hitoshi*; Fujita, Masaki*; Funakoshi, Kenichi*; Gong, W.*; et al.
Quantum Beam Science (Internet), 1(3), p.9_1 - 9_59, 2017/12
The neutron instruments suite, installed at the spallation neutron source of the Materials and Life Science Experimental Facility (MLF) at the Japan Proton Accelerator Research Complex (J-PARC), is reviewed. MLF has 23 neutron beam ports and 21 instruments are in operation for user programs or are under commissioning. A unique and challenging instrumental suite in MLF has been realized via combination of a high-performance neutron source, optimized for neutron scattering, and unique instruments using cutting-edge technologies. All instruments are/will serve in world-leading investigations in a broad range of fields, from fundamental physics to industrial applications. In this review, overviews, characteristic features, and typical applications of the individual instruments are mentioned.
Matsuba, Kenichi; Kamiyama, Kenji; Toyooka, Junichi; Zuev, V. A.*; Kolodeshnikov, A. A.*
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 11 Pages, 2017/09
To clarify jet fragmentation and accompanying cooling behaviors of molten core materials discharged into sodium, results of the out-of-pile experiments with a simulant material (AlO) were analyzed. The results clarified that while AlO jets were entirely fragmented into smaller particles during their penetration to several tenths of a meter in depth of sodium, an additional significant distance was needed to cool down the particles to the degree that thermal loading on the steel structures could be neglected. Based on the results, it is concluded that in terms of the reduction of thermal load on the lower structures in the reactor vessels, the cooling distance after fragmentation should be treated.
Ito, Daisuke*; Rivera, M. N.*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Suzuki, Toru*
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 10 Pages, 2017/09
Kawata, Ryo*; Ohara, Yohei*; Sheikh, Md. A. R.*; Liu, X.*; Matsumoto, Tatsuya*; Morita, Koji*; Guo, L.*; Kamiyama, Kenji; Suzuki, Toru
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 14 Pages, 2017/09
Nava, M.*; Ito, Daisuke*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Suzuki, Toru*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 5 Pages, 2017/07
Ito, Daisuke*; Nava, M.*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Suzuki, Toru*
Proceedings of 2017 Japan-US Seminar on Two-Phase Flow Dynamics (JUS 2017), 4 Pages, 2017/06
Yamano, Hidemasa; Suzuki, Toru; Kamiyama, Kenji; Kudo, Isamu*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 11 Pages, 2017/06
This paper describes basic visualization experiments on eutectic reaction and relocation of boron carbide (BC) and stainless steel (SS) under a high temperature condition exceeding 1500C as well as the importance of such behaviors in molten core during a core disruptive accident in a Generation-IV sodium-cooled fast reactor (750 MWe class) designed in Japan. At first, a reactivity history was calculated using an exact perturbation calculation tool taking into account expected behaviors. This calculation indicated the importance of a relocation behavior of the BC-SS eutectic because its behavior has a large uncertainty in the reactivity history. To clarify this behavior, basic experiments were carried out by visualizing the reaction of a BC pellet contacted with molten SS in a high temperature-heating furnace. The experiments have shown the eutectic reaction visualization as well as freezing and relocation of the BC-SS eutectic in upper part of the solidified test piece due to the density separation.
Kamide, Hideki; Sakamoto, Yoshihiko; Kubo, Shigenobu; Oki, Shigeo; Ohshima, Hiroyuki; Kamiyama, Kenji
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
Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures in order to strengthen safety of a fast reactor since the Great East Japan Earthquake. This paper describes the progress of design study and research and development related to safety enhancement and the severe accident countermeasures. For the purpose of strengthening of decay heat removal function, several researches have been carried out on the decay heat removal in a core disruptive accident (CDA), diversity and applicability of decay heat removal systems, and thermal hydraulic evaluation methods. In order to elucidate the behavior of molten fuel during CDA, some in-pile and out-of-pile tests has been performed by international collaboration including basic experiments. Core design was also improved from the viewpoint of preventing the occurrence of severe accident.
Imaizumi, Yuya; Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Suzuki, Toru; Emura, Yuki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04
In order to simulate the typical accident conditions of the fuel relocation phase in SFRs, the molten alloy of low melting point was discharged into a shallow water pool. The distance between the nozzle exit and the bottom plate was set to a value which was indicated to be insufficient to fragment. As the experimental result, the melt jet reached the bottom plate, and dispersed in all directions along the plate, together with the progress of fragmentation. In addition, the melt temperature on the bottom plate decreased rapidly along the radius direction. These results suggest that the fragmentation which would accompany this rapid cooling would be enhanced by the plate. This enhancement would be caused by the extension of the melt-water interface when the melt was dispersed forcibly by the plate. The solidified debris remained after the discharge showed remarkable fragmentation which was assumed to be caused by the formations of small vapor bubbles in the interface.