• 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.

• 물과 미래
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• 제23권3호
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• pp.319-328
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• 1990

2차원 하구모델의 매개변수 동정을 BF좌표계로 변환된 ADI-FDM출력 수치모형을 사용하여 수행하였다. 모형에서 기본 방정식을 동수역학식과 수송방정식이 복합된 방정식을 사용하였다. Thompson식이 지배방정식을 사각평면으로 변환 하는데 사용도이ㅓㅆ으며 Thompson의 Elliptic 격자발생기법이 BF 좌표계에서 곡선격자망을 발생하는데 사용되었다. 지배방정식에서 동정될 매개변수는 마찰계수와 분산 정수이다. 출력수치모형은 BF좌표계에서 조석평균 염도 모형이 사용되었다. 최적화 기법으로 영향계수 알고리즘을 적용하였다. 또한 Lumped 모형이 동정에 고려 되었다. 본 연구는 새로운 출력 수치모형이 하구염도 모형의 매개변수 동정에 있어 유용한가를 검토 하는데 중점을 두었다. 제안된 기법은 간단한 가상모형에 실험적인 적용을 통하여 검토되었다. 검토한 결과는 제안된 기법이 하구모형에서 매개변수 동정시 도입될 수 있음을 보여주었다.

• 대한기계학회논문집B
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• 제27권6호
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• pp.685-692
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• 2003

A numerical simulation of flame propagation in a micro combustor was carried out. Combustor has a sub -millimeter depth cylindrical internal volume and axisymmetric one-dimensional was used to simplify the geometry. Semi-empirical heat transfer model was used to account for the heat loss to the walls during the flame propagation. A detailed chemical kinetics model of $H_2/Air$ with 10 species and 16 reaction steps was used to calculate the combustion. An operator-splitting PISO scheme that is non-iterative, time-dependent, and implicit was used to solve the system of transport equations. The computation was validated for adiabatic flame propagation and showed good agreement with existing results of adiabatic flame propagation. A full simulation including the heat loss model was carried out and results were compared with measurements made at corresponding test conditions. The heat loss that adds its significance at smaller value of combust or height obviously affected the flame propagation speed as final temperature of the burnt gas inside the combustor. Also, the distribution of gas properties such as temperature and species concentration showed wide variation inside the combustor, which affected the evaluation of total work available of the gases.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1347-1360
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• 2009

A multi-dimensional component for the thermal-hydraulic system analysis code, MARS, was developed for a more realistic three-dimensional analysis of nuclear systems. A three-dimensional and two-fluid model for a two-phase flow in Cartesian and cylindrical coordinates was employed. The governing equations and physical constitutive relationships were extended from those of a one-dimensional version. The numerical solution method adopted a semi-implicit and finite-difference method based on a staggered-grid mesh and a donor-cell scheme. The relevant length scale was very coarse compared to commercial computational fluid dynamics tools. Thus a simple Prandtl's mixing length turbulence model was applied to interpret the turbulent induced momentum and energy diffusivity. Non drag interfacial forces were not considered as in the general nuclear system codes. Several conceptual cases with analytic solutions were chosen and analyzed to assess the fundamental terms. RPI air-water and UPTF 7 tests were simulated and compared to the experimental data. The simulation results for the RPI air-water two-phase flow experiment showed good agreement with the measured void fraction. The simulation results for the UPTF downcomer test 7 were compared to the experiment data and the results from other multi-dimensional system codes for the ECC delivery flow.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.439-458
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• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

• 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.

• 대한토목학회논문집
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• 제14권6호
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• pp.1385-1394
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• 1994

밀도효과(密度效果)를 고려한 물의 유동(流動)에 관한 3차원(次元) 수치모형(數値模型)을 수립하여, 정상수역(靜上水域)으로의 표면온배수(表面溫排水) 확산문제(擴散問題)를 해석하였다. 수립된 수치모형(數値模型)은 수심방향에 대해 정규화(正規化)한 좌표(座標)(${\sigma}$-coordinate)에서 무차원화(無次元化)된 식들을 사용하며, 시간(時間) 적분방법(積分方法)으로는 반음해법(半陰解法)을 사용하여 계산시간의 효율성(效率性)을 도모하였다. 온배수확산(溫排水擴散) 수리실험결과(水理實驗結果)와의 비교를 통하여 모형의 신뢰성(信賴性)을 검토하였으며, 온배수(溫排水) 확산(擴散)(밀도류(密度流)) 계산시 연직확산계수(鉛直擴散係數)와 성층효과(成層效果)를 고려하기 위해 사용되는 안정함수(安定函數)의 여러 형태에 대한 계산결과를 비교하였다. 수립된 모형은 수리실험자료(水理實驗資料)와 일치하는 양호한 계산결과를 보였다. 온배수 확산 모의시 연직(鉛直) 확산계수(擴散係數)의 공간적(空間的) 분포(分布)를 고려해야함을 확인할 수 있었으며, 표면온배수 확산을 정확히 모의하기 위해서는 기존에 널리 사용되는 안정함수(安定函數)가 수정될 필요가 있었다.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3909-3918
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• 2022

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the high-temperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.