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Tagami, Hirotaka; Tobita, Yoshiharu
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
When fuel melt occurs and it interacts with coolant in severe accidents in SFRs, it is solidified and fragmented to particles called as debris. The debris sediments and forms debris beds on structure surface. It is important to confirm whether its thickness exceeds the coolable limit to evaluate the coolability. On the other hand, because a self-leveling behavior relocates the debris and changes the bed thickness, the behavior also must be evaluated at the same time. However, no computer code to simulate this behavior exists. Therefore, this study aims at the development of computer code to simulate the self-leveling behavior on the SIMMER code. The development consists of two necessary steps. About the first step, a macroscopic model developed for fluidized bed is applied. For the second step, large deformation method is modified to be capable in multi-phase flow model. The developed code succeeded in reproducing two experiments relating to self-leveling behavior.
Shamsuzzaman, M.*; Horie, Tatsuro*; Fuke, Fusata*; Kai, Takayuki*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*; Tagami, Hirotaka; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
Investigation on sedimentation behavior of debris is important to evaluate the sequence of core disruptive accident in SFR. To clarify this behavior, a series of experiments was performed by gravity driven discharge of solid debris from a nozzle into a water pool. The discharged debris accumulates on the collector plate at the bottom, forming either a Gaussian-type convex or ring-type concave mound depending on the experiment parameters. Three types of spherical debris with three diameters are employed to study the effect of experiment parameters on mound height of debris bed. During the experiment, mound height becomes decreasing with nozzle diameter and increasing with debris volume, which exhibits descending tendency in asymmetrical fashion with density variation and an unalike variation in mound height was observed with debris diameter. An empirical model was developed applying dimensional analysis to predict the variation in mound height of debris bed during sedimentation process.
Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyoka, Junichi; Matsuba, Kenichi; Zuyev, V. A.*; Pakhnits, A. V.*; Vurim, A. D.*; Gaidaichuk, V. A.*; Kolodeshnikov, A. A.*; et al.
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12
In order to eliminate energetics potential in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner duct structure has been considered. Recently, a design option which leads molten fuel to discharge upward is considered to minimize developmental efforts for the fuel subassembly fabrication. In this paper, a series of out-of-pile tests and one in-pile test were presented. The out-of-pile tests were conducted to investigate the effects of driving pressures on upward discharge, and the in-pile test was conducted to demonstrate a sequence of upward discharge behavior of molten-fuel. Based on these experimental results, it is concluded that the most of molten-fuel is expected to complete discharging upward before core melting progression, and thereby, introduction of the fuel subassembly with the upward discharge duct has the sufficient potential to eliminate energetics events.
Hagiwara, Hiroyuki; Yamada, Yumi*; Eto, Masao*; Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
The self-actuated shutdown system (SASS), which is selected for Japan Sodium-cooled Fast Reactor (JSFR), is a passive reactor shutdown system utilizing a Curie point electromagnet (CPEM). With CPEM, an excessive fuel outlet temperature rise is sensed and the control rods are released into the core, and the reactor can be shutdown. Therefore it is important for feasibility of SASS to be established by assuring a quick response of CPEM to the coolant temperature rise. In this paper, a device named "flow collector", which collects flows discharged from six fuel subassemblies surrounding CPEM backup control rods, has been proposed to ensure a shorter response time.
Cheng, S.; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Gondai, Yoji*; Nakamura, Yuya*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
Ebara, Shinji*; Konno, Hiroaki*; Hashizume, Hidetoshi*; Kaneko, Tetsuya; Yamano, Hidemasa
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
In this study, the characteristics of pressure fluctuation in a dual elbow piping simulating the cold-leg piping of the JSFR was elucidated by conducting a pressure measurement test using a scale model. As a result of the experiment, it was clarified that the pressure fluctuation characteristics of the dual elbow flow was very similar to that of the single elbow flow in and near the first elbow.
Kubota, Ryuzaburo*; Yamada, Yumi*; Koyama, Kazuya*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kubo, Shigenobu; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
This paper describes a melt-down event progression revealed by a numerical simulation in the protected loss of heat sink (PLOHS) event for Japan Sodium-cooled Fast Reactor (JSFR). A multi-component multi-field computer code, MUTRAN, has been applied in order to simulate complicated core material motions and associated heat-transfer phenomena among the materials in a degradation core. The analyses with MUTRAN covered core degradation behaviors from the intact geometry and addressed the two initial states: one was the core without the coolant as the leakage type, and the other was the core covered by the coolant only up to the top of the fissile fuel as the boiling type. The analyses revealed representative event progression.
Ito, Kei; Ezure, Toshiki; Ohno, Shuji; Kamide, Hideki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 5 Pages, 2012/12
A free surface vortex is considered as one of important phenomena which may cause gas entrainment (GE) in sodium-cooled fast reactors. In this study, a new theoretical vortex model with realistic downward velocity distribution is proposed. This model is derived from the axisymmetric Navier-Stokes equation as well as the Burgers model, but the downward velocity distribution is considered. As the verification, the new model is applied to the evaluation of a simple vortex experiment, and shows good agreements with the experimental data in terms of the free surface shape. In addition, it is confirmed that the Burgers vortex model can gives similar results to the new vortex model when the downward velocity gradient is calculated appropriately.
Onishi, Yuki*; Takata, Takashi*; Yamaguchi, Akira*; Uchibori, Akihiro; Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
In the steam generator of sodium-cooled fast reactor (SFR), self-wastage phenomena is a crack enlargement on the heat transfer tube itself caused by sodium-water reaction, a quantification of the self-wastage phenomenon is of importance from the viewpoint of safety assessment. In this study, we propose a numerical approach to evaluate the self-wastage phenomena and investigate a crack enlargement using SERAPHIM code. In the analysis, two-dimensional initial crack is assumed based on SWAT-4 experiment. The wastage rate was estimated by Arrhenius type equation, and re-meshing arrangement was performed by cut down from a part of tube in the initial model with the wastage amount. After simulated again using the re-meshing models, the resulting SWR products were distributed not only circumferential direction but also radial direction.
Guo, L.*; Morita, Koji*; Tobita, Yoshiharu
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
In the safety analysis of liquid-metal fast reactors, thermal-hydraulic phenomena of multicomponent, multiphase flows in core disruptive accidents are regarded as particular difficulties. Accurate prediction of dispersed particle behaviors in such complicate flows is one of the key issues to be solved in numerical simulations. On the other hand, bubbling fluidization of particle beds is not only considered as an essential phenomenon in some industry areas, but also employed to understand the particle behaviors in the research field. In this study, a hybrid method for numerical simulations of bubbling fluidized beds was developed by combining the discrete element method with the multi-fluid model. A typical system of bubbling fluidized beds with glass particles is analyzed to validate the developed coupling algorithm. It was indicated that the present models and methods could provide a useful means for the numerical simulation of bubbling fluidization phenomena in particle beds.
Suzuki, Toru; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu; Nakai, Ryodai; Koyama, Kazuya*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
The JAEA has selected the advanced loop-type fast reactor JSFR as the most promising concept for the commercialization. The safety design requirements of JSFR for Design Extension Condition are the control of severe plant conditions, including the prevention of accident progression and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, the In-Vessel Retention (IVR) against Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the achievement of IVR are evaluated. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulation. It was concluded that mechanical/thermal failures of the reactor vessel could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.
Matsuba, Kenichi; Kamiyama, Kenji; Konishi, Kensuke; Toyoka, Junichi; Sato, Ikken; Zuev, V. A.*; Kolodeshnikov, A. A.*; Vasilyev, Y. S.*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12
A series of fragmentation tests (FR tests) for molten oxide was conducted to obtain experimental knowledge on the distance for fragmentation of molten core material discharged into the lower sodium plenum. Approx. 7
14 kg of molten alumina was discharged into a sodium pool (depth: 1.3 m, diameter: 0.4 m, temperature: approx. 673 K) through a duct (inner diameter: 4063 mm). The test results showed that the molten alumina was fragmented into particles within 1 m from the surface of the sodium pool. The estimated distances for fragmentation in the FR tests were roughly 6070% lower than the predictions by the existing representative correlation. The experimental knowledge confirms the possibility that the distance for fragmentation of the molten core material can be significantly reduced due to thermal interactions in the lower sodium plenum.
Naruto, Kenichi*; Kurisaka, Kenichi
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12
Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa; Sakai, Takaaki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
Yoshida, Kazuhiro*; Sakata, Hideyuki*; Sago, Hiromi*; Shiraishi, Tadashi*; Oyama, Kazuhiro*; Hagiwara, Hiroyuki*; Yamano, Hidemasa; Yamamoto, Tomohiko
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12
To prevent the vortex cavitations, asymmetric flow in the upper plenum due to the radial slit with upper internal structure (UIS) has been mitigated by installing a cylindrical structure named as dummy plug instead of the fuel handling machine only used for refueling period. In this study, the extended brim and the division plate at the slit of UIS have been proposed in order to improve flow pattern in upper plenum for the purpose of the vortex cavitation prevention.
Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Matsumoto, Tatsuya*; Morita, Koji*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
In this study, numerical calculations were carried out, providing that the fluid-dynamics model of SIMMER-IV was valid to simulate the sloshing behavior. Comparing to the conventional two-dimensional simulation, it was found that the three-dimensional simulation can mitigate the fuel compaction to the center because the effect of circumferential momentum dissipation can be addressed. From these calculations, the validity of the SIMMER-IV code was confirmed for the sloshing behavior.
Hayakawa, Satoshi*; Ishikura, Shuichi*; Watanabe, Osamu*; Kaneko, Tetsuya*; Yamano, Hidemasa; Tanaka, Masaaki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
The present methodology was applied to the analysis for the 1/3-scale experiment of the hot-leg pipe of JSFR, and the predicted stress values were compared with the measured stress values. The predicted stress values were underestimated in the case of using the intact pressure fluctuations obtained by the unsteady fluid flow analysis. Therefore, the improvement of the prediction accuracy of the pressure fluctuations on the pipe wall was attempted.
Ohno, Shuji; Hamase, Erina; Kamide, Hideki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 5 Pages, 2012/12
Sensitivity analyses of sodium leak and fire are performed to identify the dominant factors for the accident consequences. The analyses with a multi-cell zone model code SPHINCS treat sodium spray and pool simultaneous combustion and heat-mass transfer behaviors in a large-scale two-cell geometry. Atmospheric gas pressure increase and temperature increase of floor steel plate below the sodium pool are analyzed as the figures of merit to be directly focused on. The analyses clarify the important and dominant factors of the phenomena in the accident sequence quantitatively, resulting in the acquirement of the knowledge to conduct the appropriate code validation activity and to discuss the uncertainty in the safety evaluation results.
Fujita, Kaoru; Yamano, Hidemasa; Kubo, Shigenobu*; Eto, Masao*; Yamada, Yumi*; Toyoshi, Akira*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 6 Pages, 2012/12
no abstracts in English
Shen, X.*; Nakamura, Hideo
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12