• 에너지공학
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• 제22권1호
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• pp.8-16
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• 2013

A core catcher is considered as a promising engineered system to stabilize the molten corium in the containment during a postulated severe accident in a nuclear power plant. Conceptually, the core catcher consists of a carbon steel body, sacrificial material, protection material, and engineered cooling channel. The cooling capacity of the engineered cooling channel should be guaranteed to remove the decay heat of the molten corium. The flow in ex-vessel core catcher is a combined problem of a two-phase flow in the engineered cooling channel and a single-phase natural circulation in the whole core catcher system. In this study, the analysis of the test facility for the core catcher using the CUPID code, which is a three-dimensional thermal-hydraulic code for the simulation of two-phase flows, was carried out to evaluate its cooling capacity.

• 한국전산유체공학회지
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• 제15권2호
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• pp.86-94
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• 2010

A thermal-hydraulic code, named CUPID, has been developed for the analysis of transient two-phase flows in nuclear reactor components. A two-fluid three-field model was used for steam-water two-phase flows. To obtain numerical solutions, the finite volume method was applied over unstructured cell-centered meshes. In steam-water two-phase flows, a phase change, i.e., evaporation or condensation, results in a great change in the flow field because of substantial density difference between liquid and vapor phases. Thus, two-phase flows are very sensitive to the local pressure distribution that determines the phase change. This in turn puts emphasis on the accurate evaluation of local pressure gradient. This paper presents a new reconstruction method to evaluate the pressure gradient at cell centers on unstructured meshes. The results of the new scheme for a simple test function, a gravity-driven cavity, and a wall boiling two-phase flow are compared with those of the previous schemes in the CUPID code.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2010년 춘계학술대회논문집
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• pp.380-388
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• 2010

A thermal-hydraulic code, named CUPID, has been developed for the analysis of transient two-phase flows in nuclear reactor components. A two-fluid three-field model was used for steam-water two-phase flows. To obtain numerical solutions, the finite volume method was applied over unstructured cell-centered meshes. In steam-water two-phase flows, a phase change, i.e., evaporation of condensation, results in a great change in the flow field because of substantial density difference between liquid and vapor phases. Thus, two-phase flows are very sensitive to the local pressure that determines the phase change. This in turn puts emphasis on the accurate evaluation of local pressure gradient. This paper presents a new numerical scheme to evaluate the pressure gradient at cell centers on unstructured meshes. The results of the new scheme for a simple test function a gravity-driven cavity, and a wall boiling two-phase flow are compared with those of the previous schemes in the cupid code.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1400-1409
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• 2023

In this study, three condensation models of the CUPID code, i.e., the resolved boundary layer approach (RBLA), heat and mass transfer analogy (HMTA) model, and an empirical correlation, were tested and validated against the COPAIN and CAU tests. An improvement on HMTA model was also made to use well-known heat transfer correlations and to take geometrical effect into consideration. The RBLA was a best option for simulating the COPAIN test, having mean relative error (MRE) about 0.072, followed by the modified HMTA model (MRE about 0.18). On the other hand, benchmark against CAU test (under natural convection and occurred on a slender tube) indicated that the modified HMTA model had better accuracy (MRE about 0.149) than the RBLA (MRE about 0.314). The HMTA model with wall function and the empirical correlation underestimated significantly, having MRE about 0.787 and 0.55 respectively. When using the HMTA model, consideration of geometrical effect such as tube curvature was essential; ignoring such effect leads to significant underestimation. The HMTA and the empirical correlation required significantly less computational resources than the RBLA model. Considering that the HMTA model was reasonable accurate, it may be preferable for large-scale simulations of containment.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2276-2296
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• 2022

This study validates the applicability of the CUPID code for simulating subcooled wall boiling under high-pressure conditions against number of DEBORA tests. In addition, a new numerical technique in which the interfacial momentum non-drag forces are calculated at the cell faces rather than the center is presented. This method reduced the numerical instability often triggered by calculating these terms at the cell center. Simulation results showed good agreement against the experimental data except for the bubble sizes in the bulk. Thus, a new model to calculate the Sauter mean diameter is proposed. Next, the effect of the relationship between the bubble departure diameter (D_dep) and the nucleation site density (N) on the performance of the Wall Heat Flux Partitioning (WHFP) model is investigated. Three correlations for D_dep and two for N are grouped into six combinations. Results by the different combinations show that despite the significant difference in the calculated D_dep, most combinations reasonably predict vapor distribution and liquid temperature. Analysis of the axial propagations of wall boiling parameters shows that the N term stabilizes the inconsistences in D_dep values by following a behavior reflective of D_dep to keep the total energy balance. Moreover, ratio of the heat flux components vary widely along the flow depending on the combinations. These results suggest that separate validation of D_dep correlations may be insufficient since its performance relies on the accompanying N correlations.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2009년 춘계학술대회논문집
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• pp.290-297
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• 2009

가압 경수로의 주요 기기에서 발생할 수 있는 과도 2상 유동(Two-phase flow) 현상에 대한 해석을 수행하기 위해 원자로 기기 열수력 해석 코드를 개발 중에 있다. 개발 중인 기기 열수력 해석 코드는 지배 방정식으로 Two-phase, three-field model을 사용하고 있으며, 복잡한 기하학적 형상의 원자로 기기를 모사하기 위해 비정렬 격자계(Unstructured grid)를 활용하고 있다. 수치해석 기법으로는, 원자로 계통 해석코드 RELAP5가 사용 중이며 대부분의 원자로 내 2상 유동 조건에서 안정적이며 정확하다고 알려진 Semi-implicit 방법을 적용하였다. 그러나 기존의 Semi-implicit 방법은 1차원, 엇갈림격자(Staggered grid)에 대해 개발되었기 때문에, 이를 다차원, 비정렬, 비엇갈림 격자(Non-staggered grid)에 적용하기 위해 기존의 Semi-implicit 방법을 수정하였다. 본 논문에서는 Semi-implicit 방법의 대류항을 이차정확도를 갖도록 확장하였으며, 이차정확도에 의한 수치확산의 감소를 평가하기 위해 수행된 수치시험의 결과를 기술하였다. 이차정확도 및 일차정확도로 계산된 값을 해석해 또는 격자 수렴성 시험을 통해 평가해 본 결과, 이차정확도 계산시 수치 확산의 감소 확인하였다.

• 한국전산유체공학회지
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• 제18권2호
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• pp.99-108
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• 2013

The standard drag force and virtual mass force, which exert to the primary flow direction, are generally considered in two-phase analysis computational codes. In this paper, the lift force, wall lubrication force, and turbulent dispersion force including turbulence models, which are essential for a computational multi-fluid dynamics model and play an important role in motion perpendicular to the primary flow direction, were introduced and verified with conceptual problems.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.