Two-phase Flow Analysis Code Research Articles

Numerical simulation of two-phase flow in 4x4 simulated bundle

An evaluation methodology of a thermal-hydraulics based on a mechanism in light water reactors (LWRs) is needed from a viewpoint of the safety analysis during normal operation and unanticipated transient such as under a severe accident. Currently, the evaluation of safety for the nuclear reactor has been implemented by a best estimate (BE) code and subchannel analysis code. These analysis codes contain models and empirical correlations. Therefore, the full-scale mock-up tests are needed to evaluate the reliability and validation of code. And the model and empirical correlation are allowed to be applied only in the range where the experiments were implemented. The large mock-up tests are once again needed in order to consider the new geometry and boundary conditions when the design of components is changed. Hence, the 3D detailed numerical simulation by the mechanistically based method is expected to be applied for the preliminary analysis to improve the design of fuel assemblies and evaluate the safety. This 3D detailed numerical simulation can reduce the large mock-up tests. The detailed numerical simulation method can provide much information relating to the two-phase flow such as the bubble size, its velocity, and detailed void distribution which, for example, are needed to predict the critical heat flux based on the mechanism. Moreover, JAEA is implementing the development of the nuclear-thermal-coupling code by using a detailed two-phase flow analysis code based on the VOF method like a JUPITER code. In this study, the numerical simulation of two-phase flow in the 4x4 bundle was examined by numerical simulation code JUPITER in order to examine the possibility of the JUPITER code for the large scale two-phase flow analysis. The simulation results are verified by the previous experimental data of two-phase flow.

Multi-scale analysis of an ATLAS-MSLB test using the coupled CUPID/MARS code

This study describes a multi-scale nuclear safety analysis method for a main steam line break accident, which focuses on the 3-dimensional approach for the coolant temperature in the reactor pressure vessel related to the re-criticality and/or pressurizer thermal shock. The direct coupled code, CUPID/MARS, where CUPID is a 3-dimensional two-phase flow analysis code and MARS is a 1-dimensional nuclear system analysis code, was validated against a main steam line break test in the ATLAS facilities, which are a 1/2 scale integral test loop for a Korean APR1400 PWR. The calculation indicates that the suggested 4-step method, which expands the previous 2-step method of steady-state and transient calculations of a 1-dimensional system code, is valid for analyzing a direct 1-D/3-D coupled safety analysis, and the 3-dimensional temperature distribution in the downcomer agrees with the measured result.

Numerical approach to siphon break phenomena in a research reactor pool using the CUPID code

Pool-type reactors, as research reactors, are equipped with a siphon break line to prevent continuous siphonage flow that can cause core uncovery during a loss of coolant accident (LOCA). In this study, the siphonage flow rate and siphon break phenomena during a LOCA in a pool-type research reactor are assessed using the three-dimensional two-phase transient flow analysis code, CUPID. The calculation indicates that the siphonage LOCA break flow rate relies on the wall friction model for the main drainage pipe, and the initiation of a siphon break is significantly affected by the interfacial drag model. The CUPID simulation of the siphon break confirms that the siphonage LOCA break flow rate and the initiation of siphon break phenomena are accurately predicted using the three-dimensional two-phase flow analysis code with the relevant physical models for wall friction and interfacial drag.

An implicit numerical scheme for the simulation of three-dimensional two-phase flows in light water nuclear reactors

ABSTRACTAn implicit numerical scheme is presented for solving the two-fluid model widely used in the analysis of a gas–liquid two-phase flow in light water nuclear reactors (LWRs). The pressure equation is established by combining the momentum and mass conservation equations. The implicit calculation of each governing equation is separated into phase- and space-link steps. In the phase-link step, the interfacial momentum and heat transfers are implicitly calculated. Then the solution accounting for the convection and diffusion terms is calculated simultaneously in space. The phase- and space-link steps are repeated for convergence. The numerical scheme is implemented in CUPID, which is a multidimensional two-phase flow analysis code for LWRs, and verified against a set of conceptual two-phase flow problems which include typical thermal hydraulic phenomena in LWRs. Calculations are performed using four numerical schemes, semi-implicit ICE and SMAC schemes, an implicit scheme, and an implicit scheme with increased time step size, and the results are discussed.

F232 二相流解析コードTPFITの垂直矩形管における空気と水二相流に対する検証(OS9 熱・流動(6))

In Japan Atomic Energy Agency, the detailed two-phase flow analysis code TPFIT has been developed to simulate and evaluate two-phase flow characteristics in nuclear systems. In this study, a numerical simulation of bubbly flow in a vertical square channel was performed to validate the applicability of TPFIT code on bubbly flow simulation. By checking bubble distribution development in the flow direction, the calculation of the forces acting on bubbles was validated through comparing simulation results and experimental results from Matos et al. (2004). Comparisons between the experimental and numerical data revealed, in general, good agreement except serious bubble coalescence appeared in numerical simulation.

Numerical study for CANDU moderator temperature prediction by using the two-phase flow analysis code, CUPID

KAERI has been developing a component-scale thermal–hydraulics code, CUPID. The code adopts a three-dimensional, transient, three-field model for two-phase flow. In this study, we investigated the thermal hydraulic behavior of the moderator inside the Calandria tank of a CANDU reactor by using the CUPID code. At first, we have validated the CUPID code using the experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational geometry around the Calandria tube bundles, a porous media approach was applied for that region and the flow resistance inside the porous media zone was modeled by an empirical correlation. An open media is applied to generate the outer fluid layer including the inlet nozzles. Computational grids near the inlet nozzles should be well-generated because the flow field is very sensitive to the momentum flux from the nozzle. Since the axial flow can be assumed to be invariant for this experiment, a two-dimensional approach was adopted. The mixed flow pattern of forced and natural convection inside the Calandria vessel has been successfully predicted by the CUPID code. The analysis has been further extended to two-phase flow conditions and, then, a map of the local maximum moderator temperature in the Calandria vessel versus the injection flow rate was derived, which can be used to predict the local subcooled margin in the vessel.

303 界面追跡法による垂直管内上昇気泡流の数値解析(OS9 オーガナイズドセッション《熱・流体・エネルギー変換工学の進展》)

In Japan Atomic Energy Agency, the detailed two-phase flow analysis code TPFIT has been developed to simulate and evaluate two-phase flow characteristics in nuclear systems [1]. In this study, a numerical simulation of bubbly flow in a vertical circular pipe was performed to validate the applicability of TPFIT code on bubbly flow simulation. By checking bubble distribution development in the flow direction, the calculation of the forces acting on bubbles was validated through comparing simulation results and experimental results from Lucas et al. (2005), who used wire-mesh sensor to obtain a high resolution of the gas fraction data in space as well as in time.

2상 유동 해석코드 CUPID를 이용한 CANDU 원자로 감속재 열수력 예비해석

본 연구에서는 기기 스케일 2상 유동(Two-phase flow) 해석 코드 CUPID를 사용하여 CANDU 원자로의 칼란드리아 용기 내부 감속재의 열수력 거동을 분석하기 위한 사전연구를 수행하였다. 먼저, Stern 연구소에서 수행한 단상유동 실험 3종류를 이용하여 CUPID 코드를 검증하였다. 칼란드리아 관다발 영역 격자생성의 복잡성을 피하기 위하여 다공성 매질 모델을 해당 영역에 적용하였고, 다공성 매질 영역의 유동 저항은 실험에서 얻은 관계식을 이용하여 계산하도록 하였다. 계산결과, CUPID 코드는 칼란드리아 용기 내부의 강제 및 자연 대류의 혼합 유동 양식을 성공적으로 예측하였다. 다음으로 2상 유동이 발생하는 경우를 해석하였다. 이들 계산을 통해 CUPID 코드의 CANDU 원자로 감속재 해석 능력을 보였다. 또한, 국부 과냉각 여유도를 예측하는데 사용할 수 있는 유입유량 대비 칼란드리아 용기의 국부 최대 감속재 온도 그래프를 제시하였다. A transient, three-dimensional, two-phase flow analysis code, CUPID, has been developed in KAERI. In this work, we performed a preliminary analysis using the CUPID code to investigate the thermal-hydraulic behavior of the moderator in the Calandria vessel of a CANDU reactor. At first, we validated the CUPID code using the three experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational mesh around the Calandria tube bundles, a porous media approach was applied for the region. The pressure drop in the porous media zone was modeled by an empirical correlation. The results of the calculations showed that the CUPID code can predict the mixed flow pattern of forced and natural convection inside the Calandria vessel very well. Thereafter, the analysis was extended to a two-phase flow condition. Also, the local maximum temperature in the Calandria vessel was plotted as a function of the injection flow rate, which may be utilized to predict the local subcooling margin.

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.

Implementation of a second-order upwind method in a semi-implicit two-phase flow code on unstructured meshes

A two-phase flow analysis code, CUPID, has been developed for a realistic simulation of thermal–hydraulic phenomena in nuclear reactor components. In the CUPID code, a two-fluid three-field model is adopted and the governing equations are solved on unstructured meshes. To obtain the numerical solution, the semi-implicit method of the RELAP5 code was used with some modifications for a cell-centered finite volume method. In this work, a second-order upwind method was implemented for the convective terms of the CUPID code. To get the slopes of the convective quantities, we adopted the Frink’s reconstruction method (1994, AIAA Paper 94-0061) and modified it for an application to arbitrary polyhedral cells. To stabilize the numerical solutions, the Barth and Jesperson’s slope limiter was used. In order to evaluate the enhanced accuracy and ensure the robustness of the implemented scheme, numerical tests were performed using conceptual single- and two-phase flow problems, which include a strong phase change and very heterogeneous phase distributions.

20503 稠密炉心内3次元気泡流構造に関する大規模シミュレーション(大規模数値解析(1),OS11 大規模数値解析)

In order to predict the water-vapor two-phase flow structure in a fuel bundle of an advanced light water reactor, a large-scale numerical simulation was carried out using a newly developed two-phase flow analysis code and a highly parallel vector supercomputer. Conventional analysis methods with a two-fluid model need composition equations and empirical correlations based on the experimental data. Therefore, it is difficult to obtain high prediction accuracy when experimental data are nothing. Then, a new two-phase flow analysis method based on the latest computational science was proposed. This paper describes the predicted bubbly flow dynamics in three-dimensional minichannels and the predicted void fractions in the simulated tight-lattice fuel channels. Regarding the coalescence and fragmentation of bubbles the useful knowledge was obtained, and also the high prospect was acquired on the possibility of development of the thermal design procedure of the advanced nuclear reactors with large-scale simulations.

Numerical Analysis of a Water-Vapor Two-Phase Film Flow in a Narrow Coolant Channel with a Three-Dimensional Rectangular Rib

A reduced-moderation light water reactor (RMWR), which is studied by the Japan Atomic Energy Research Institute, needs to the volume ratio between the coolant and fuel less than 0.2, in order to achieve 1.1 or more high breeding ratios. The RMWR fuel assembly consists of many fuel rods with a diameter of around 10mm with the triangular tight-lattice configuration in the radial direction. A minimum gap width between adjacent fuel rods is about 1mm. Moreover, spacers are set to keep uniformly the gap width between fuel rods horizontally. These spacers are installed at a fixed interval to the flow direction in a reactor core. As a result of this, a complicated flow is formed around each spacer because of a large flow disturbance. Then, fundamental liquid film flow behavior in a narrow channel with a three-dimensional rectangular rib was analyzed numerically using a newly developed two-phase flow analysis code. From this study the following conclusions were derived: a wake behind the spacer rib takes place easier when the initial film thickness is short and then the gas velocity is higher than the liquid velocity; and, the liquid film thickness reduces with increasing the gas velocity.

狭隘流路内の大規模気泡流挙動に関する数値シミュレーション(S19-3 混相流の多次元ビジュアリゼーション(気液・気泡), S19 混相流の多次元ビジュアリゼーション)

In order to predict the water-vapor two-phase flow structure in a fuel bundle of an advanced light water reactor, a large-scale numerical simulation was carried out using a newly developed two-phase flow analysis code and a highly parallel vector supercomputer. Conventional analysis methods such as subchannel codes need composition equations based on the experimental data. In case that there are no experimental data for the advanced light water reactor, therefore, it is difficult to obtain high prediction accuracy. Then, a new analysis method using the large-scale direct numerical simulation of water-vapor two-phase flow was proposed. This paper, describes the predicted results of a large-scale bubbly flow in a narrow square duct. Regarding the mechanism of bubble coalescence and fragmentation the useful knowledge was obtained.

C212 障害物を伴う平行平板流路内気液二相流の界面挙動に関する数値予測

Three-dimensional computations were carried out using a newly developed two-phase flow analysis code in order to analyze numerically free surface characteristics between water and vapor in narrow fuel channels with rectangular spacer ribs for reduced-moderation light water reactors. Calculations were conducted under isothermal condition. Major calculation parameters are water and vapor velocities at the channel inlet and initial thickness of water on the channel floor. The Liquid film flow behavior around one or two spacer ribs was predicted quantitatively.

Development of Two-Phase Flow Analysis Code by 2V2T Model, (I)

For development of a two-phase flow analysis code by use of two-velocity two-temperature (2V2T) model, six basic equations completely independent and consistent with one another are derived for several possible types of thermodynamical unequilibrium conditions. Characteristic of the basic equations is that evaporation and condensation take place in the saturated water and saturated vapor separately. These phase change equations in the saturated state are rigorously derived using the thermodynamical law. The energy conservation equation of each phase is derived from the well known total two-phase flow energy equation, using the first law of thermodynamics and conservation equations of mass and momentum of each phase. This derivation method will give assurance that all conservation equations are consistent with one another. To form simplest 2V2T model, the terms of wall and interphase friction, and heat transfer to the two-phase flow and between phases are considered in this paper.

Development of two-phase flow analysis code by 2V2T model, (I). Derivation of basic equations.

For development of a two-phase flow analysis code by use of two-velocity two-temperature (2V2T) model, six basic equations completely independent and consistent with one another are derived for several possible types of thermodynamical unequilibrium conditions. Characteristic of the basic equations is that evaporation and condensation take place in the saturated water and saturated vapor separately. These phase change equations in the saturated state are rigorously derived using the thermodynamical law. The energy conservation equation of each phase is derived from the well known total two-phase flow energy equation, using the first law of thermodynamics and conservation equations of mass and momentum of each phase. This derivation method will give assurance that all conservation equations are consistent with one another. To form simplest 2V2T model, the terms of wall and interphase friction, and heat transfer to the two-phase flow and between phases are considered in this paper.