Advanced Interface Tracking Method Research Articles

Three-Dimensional Simulation of Vapor Bubble Growth in Superheated Water Due to the Convective Action by an Interface Tracking Method

Abstract In this paper, the growth of a rising vapor bubble in superheated water was numerically studied using an advanced interface tracking method, called the intersection marker (ISM) method. The ISM method is a hybrid Lagrangian–Eulerian front-tracking algorithm that can model an arbitrary three-dimensional (3D) surface within an array of cubic control volumes (CCV). The ISM method has cell-by-cell remeshing capability that is volume conservative, maintains surface continuity, and is suited for tracking interface deformation in multiphase flow simulations. This method was previously used in adiabatic bubble rise simulation with no heat and mass transfers to or from the bubble were considered. This work will extend the ISM method's application to simulate vapor bubble growth in superheated water with the inclusion of additional physics, such as the convective heat transfer mechanism and the phase-change. Coupled with an in-house variable-density and variable-viscosity single-fluid flow solver, the method was used to simulate vapor bubble growth due to the convective action. The forces such as the surface tension and the buoyancy were included in the momentum equation. The source terms for the mass transfer were also modeled in the computational fluid dynamics governing equations to simulate the growth. Bubble properties such as size, shape, velocity, drag coefficient, and convective heat transfer coefficient were predicted. Effects of surface tension and temperature on the bubble characteristic were also discussed. Obtained numerical results were compared against the analytical and past works and found to be in good agreement.

E112 詳細二相流解析コードTPFITの水噴流に対する検証(II) : 高流量に対する境界条件の検討(OS9 熱・流動(解析))

The Japan Atomic Energy Agency has developed TPFIT, a detailed two-phase flow analysis code with an advanced interface tracking method, as part of development of a new method of thermal hydraulic design for nuclear reactor cores based on large-scale numerical simulation. In past studies, most attempts to validate TPFIT have been performed under bubbly, slug and churn flow conditions. However, to apply TPFIT to a two-phase flow in fuel assemblies, TPFIT must also be validated for annular and dispersed flow. In this study, TPFIT was applied to a water jet experiment as a first step in validating TPFIT for annular and dispersed flow conditions. The liquid film flowing on fuel rods in an annular two-phase flow was imitated as a water jet. To simulate jet, boundary conditions are important especially for high velocity conditions. In this paper, appropriate boundary condition for high flow rate (high velocity) simulations was evaluated.

10506 改良界面追跡法による燃料集合体内流路閉塞効果実験解析(産学における混相研究(2),OS.13 産学における混相研究)

10506 改良界面追跡法による燃料集合体内流路閉塞効果実験解析(産学における混相研究(2),OS.13 産学における混相研究)

Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: III - Numerical Evaluation of Fluid Mixing Phenomena using Advanced Interface-Tracking Method -

Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. However, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. We tried to verify the TPFIT code by comparing it with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.

Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology.In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low.

Numerical Simulation of Fluid Mixing Phenomena in Boiling Water Reactor Core Using Advanced Interface-Tracking Method

Thermal-hydraulic design of the current BWR is performed by correlations with empirical results of actual-size tests. Then, when the reactor of new design is developed, an actual size test is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for BWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method : TPFIT is developed. In this paper, the TPFIT code was applied to simulation of two-phase flow in modeled 2 subchannels of BWRs rod bundle, and the existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data.

Numerical Investigation of Cross Flow Phenomena in a Tight-Lattice Rod Bundle Using Advanced Interface Tracking Method

In relation to the design of an innovative FLexible-fuel-cycle Water Reactor (FLWR), investigation of thermal-hydraulic performance in tight-lattice rod bundles of the FLWR is being carried out at Japan Atomic Energy Agency (JAEA). The FLWR core adopts a tight triangular lattice arrangement with about 1 mm gap clearance between adjacent fuel rods. In view of importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study presents a statistical evaluation of numerical simulation results obtained by a detailed two-phase flow simulation code, TPFIT, which employs an advanced interface tracking method. In order to clarify mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is applied to simulate water-steam two-phase flow in two modeled subchannels. Attention is focused on instantaneous fluctuation characteristics of cross flow. With the calculation of correlation coefficients between differential pressure and gas/liquid mixing coefficients, time scales of cross flow are evaluated, and effects of mixing section length, flow pattern and gap spacing on correlation coefficients are investigated. Differences in mechanism between gas and liquid cross flows are pointed out.

OS10-10 改良界面追跡法によるBWR燃料集合体内流体混合量の評価(OS10 混相流の計測技術と解析,循環型社会における動力エネルギー技術)

Thermal-hydraulic design of the current BWR is performed by correlations with empirical results of actual-size tests. Then, when the reactor of new design is developed, an actual size test is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear rectors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for BWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method: TPFIT is developed. In this paper, the TPFIT code was applied to simulation of two-phase flow in modeled 2 subchannels of BWRs rod bundle, and the existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data.

ICONE15-10532 Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles:IV : Numerical Evaluation of Fluid Mixing Phenomena using Advanced Interface-Tracking Method

Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test that simulates its design is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear rectors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method: TPFIT is developed to get the detailed information of the two-phase flow. We tried to verify the TPFIT code comparing with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.

0613 改良界面追跡法によるBWR炉心内流体混合現象評価手法開発の現状(S45-3 原子炉システムおよびその要素技術(3),S45 原子炉システムおよびその要素技術)

0613 改良界面追跡法によるBWR炉心内流体混合現象評価手法開発の現状(S45-3 原子炉システムおよびその要素技術(3),S45 原子炉システムおよびその要素技術)

1187. 大規模シミュレーションによる稠密炉心内気液二相流特性の解明,(III)

The reduced-moderation water reactor (RMWR) core adopts a hexagonal tight-lattice arrangement with about 1mm 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 thermalhydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing an advanced interface tracking method to improve the conservation of volume of fluid.The present paper describes a part of the verification works of the two-phase flow simulation code TPFIT. The numerical results applied to liquid film falling down on an inclined flat plate are shown and compared with existing experimental results. In the results of numerical simulation, the development of the wave structures could be reproduced by the numerical simulation. The predicted average local film thicknesses almost agreed with Nusselt's mean film thickness. Moreover, the probability density function of local film thickness agreed well with the experimental result. In addition, the trends related to the film velocities were consistent with the theoretical results. From these results, it was confirmed that the TPFIT code is applicable to predict the liquid film flow behavior.