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Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, 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
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.
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.
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.
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; Toyooka, 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
Ono, Ayako; Kobayashi, Jun; Kamide, Hideki; Watanabe, Osamu*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor. The DHRS of JSFR consists of one unit of DRACS and two units of PRACS. In this study, the sodium experiments were conducted using a sodium test loop PLANDTL in order to investigate the effect of operation mode on transient behavior of thermal hydraulic in PRACS loop. The experimental results revealed the effect of increasing heat removal capacity of PRACS and the forced flow operation in PRACS loop on the thermal transient and natural circulation behavior in PRACS loop.
Ezure, Toshiki; Ito, Kei; Onojima, Takamitsu; Kimura, Nobuyuki; Kamide, Hideki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12
In this study, water experiments were performed in the 1/22 scaled upper plenum model of JSFR. Occurrence behavior of vortex cavitation was grasped quantitatively by means of the visualization and image analyses under several conditions of kinematic viscosity 
) and pressure (
). The experimental results showed that the vortex cavitation has dependence on the variation of and P. The increase of at least in the present small model, leaded to the restriction of cavitation as assumed by Burgers model. And also, the restriction level of vortex cavitation according to the increase was smaller than the evaluation using cavitation factor.
Eto, Kenichiro*; Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Ito, Kei
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12
When inert gas is present in the primary coolant of a sodium cooled fast reactor (SFR), it may cause a core power fluctuation or a reduction of heat transfer at an intermediate heat exchanger (IHX). Therefore, it is necessary to clarify an allowance level of the gas in the system design of SFR. In Japanese Sodium Fast Reactor (JSFR), a dipped plate (D/P) will be installed at the upper plenum so as to suppress a fluid fluctuation at the free surface and entrainment of argon (Ar) gas bubbles into the piping system. In the present study, an influence of the D/P on the gas behavior has been investigated using the VIBUL code. As a result, it is demonstrated that the entrainment of Ar gas is suppressed considerably by the D/P although the background void fraction, in which no Ar entrainment from the free surface is taken into account, increases comparing with that without D/P. The quantification of the allowance level of Ar entrainment is also investigated based on the computation result.
Konaka, Yuji*; Takata, Takashi*; Yamaguchi, Akira*; Ito, Kei; Ohshima, Hiroyuki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12
Bubbles of inert gases exist in a primary coolant system of sodium-cooled fast reactor. Those bubbles may cause increase of reactivity in the core and cavitation erosion and so on. Therefore it is necessary from the viewpoint of reactor safety to understand the dynamic behavior of the bubbles. In this study, the model of gas transport at the entrance nozzle has been developed using a non-dimensional correlation based on a three-dimensional analysis of dynamic bubbles behavior. As a result, it has been demonstrated that the bubble behavior is influenced by the sodium flow field, bubble radius and the shape of the entrance nozzle. The authors have also developed a correlation for the gas behavior based on the dimensionless number which corresponds to the geometric, the buoyancy and the drag force effects acting on the bubble.
Yoshida, Hiroyuki; Nagatake, Taku; Takase, Kazuyuki; Kaneko, Akiko*; Monji, Hideaki*; Abe, Yutaka*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12
Misawa, Takeharu; Takase, Kazuyuki; Mori, Hideo*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 6 Pages, 2012/12
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.