Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11
Horiguchi, Naoki; Yoshida, Hiroyuki; Abe, Yutaka*
Nuclear Engineering and Design, 310, p.580 - 586, 2016/12
Kato, Yuki; Yoshida, Hiroyuki; Yokoyama, Ryotaro*; Kanagawa, Tetsuya*; Kaneko, Akiko*; Monji, Hideaki*; Abe, Yutaka*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
Yoshida, Hiroyuki; Nagatake, Taku; Takase, Kazuyuki; Kaneko, Akiko*; Monji, Hideaki*; Abe, Yutaka*
Mechanical Engineering Journal (Internet), 1(4), p.TEP0025_1 - TEP0025_11, 2014/08
Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime
Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04
The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.
Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime
Proceedings of Japan-US Seminar on Two-Phase Flow Dynamics, p.317 - 325, 2004/12
We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with Power Company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration. In this paper, we will show the R&D plan and describe the current status on experimental and analytical studies. We will confirm the thermal-hydraulic performance in the tight-lattice bundles by this project and develop a predictable technology for the RMWR in future.
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime
Nippon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.233 - 241, 2004/09
When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1 mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.
Ito, Kei; Kunugi, Tomoaki*; Ohshima, Hiroyuki
no journal, ,
In this study, the authors develop a new manufactured solution with dynamic interfacial deformation due to a vortex, which is a simplified model of the gas entrainment behavior in a nuclear reactor. The manufactured solution is considered on an axisymmetric system and radial, circumferential and axial velocities and pressure are formulated to satisfy the continuity equation and the boundary condition on an interface. The interfacial dent grows with time and a gas bubble is generated when the lower part of the interfacial dent is pinched off. A preliminary simulation is performed on a coarse mesh to investigate the dynamic interfacial deformation on the velocity field given by the manufactured solution. As a result, a reasonable interfacial shape is simulated at each elapsed time, which implies the developed manufactured solution is a good problem to verify an interface-tracking method.
Ito, Kei; Kunugi, Tomoaki*; Koizumi, Yasuo*; Ohno, Shuji; Ohshima, Hiroyuki
no journal, ,
One thermal-hydraulics issue in a sodium-cooled fast reactor is gas entrainment (GE) phenomena. The authors have developed a numerical simulation code for gas-liquid two-phase flows with a high-precision interface-tracking method. This simulation code employs an unstructured mesh scheme to model structural geometries accurately. As for the interface-tracking method, the authors have developed an innovative method based on the high-precision volume-of-fluid method. As a result of fundamental validations, the developed code shows superior simulation accuracy to conventional codes. For example, with the developed second-order interface gradient calculation method, the developed code provides 20-30% smaller simulation error than the original PLIC method for the slotted-disk revolution problem. The GE phenomena itself are also simulated with the developed code and the quantitative agreements of entrained gas flow rates are obtained between the simulation results are the experimental data.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
no journal, ,
An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. JAEA is embarking on the establishment of methodology to simulate the two-phase flow in the fuel bundles which applies to prediction CHF. In this study, the numerical simulation of two-phase flow in 44 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. It was confirmed that the JUPITER enables to simulate the two-phase flow in the 44 simulated fuel bundle which is large calculation domain over 3 m.