• 한국전산유체공학회지
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• 제14권4호
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• pp.13-22
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• 2009

A two-phase (gas and liquid) flow analysis solver, named CUPID, has been developed for a realistic simulation of transient two-phase flows in light water nuclear reactor components. In the CUPID solver, a two-fluid three-field model is adopted and the governing equations are solved on unstructured grids for flow analyses in complicated geometries. For the numerical solution scheme, the semi-implicit method of the RELAP5 code, which has been proved to be very stable and accurate for most practical applications of nuclear thermal hydraulics, was used with some modifications for an application to unstructured non-staggered grids. This paper is concerned with the effects of interpolation schemes on the simulation of two-phase flows. In order to stabilize a numerical solution and assure a high numerical accuracy, the second-order upwind scheme is implemented into the CUPID code in the present paper. Some numerical tests have been performed with the implemented scheme and the comparison results between the second-order and first-order upwind schemes are introduced in the present paper. The comparison results among the two interpolation schemes and either the exact solutions or the mesh convergence studies showed the reduced numerical diffusion with the second-order scheme.

• 시스템엔지니어링학술지
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• 제19권2호
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• pp.46-58
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• 2023

Deterministic safety analysis is a crucial part of safety assessment, particularly when it comes to demonstrating the safety of nuclear power plant designs. The traditional approach to deterministic safety analysis models is to model the nuclear core using point kinetics. However, this simplified approach does not fully reflect the real core behavior with proper moderator and fuel reactivity feedbacks during the transient. The use of Multi-Physics approach allows more precise simulation reflecting the inherent three-dimensionality (3D) of the problem by representing the detailed 3D core, with instantaneous updates of feedback mechanisms due to changes of important reactivity parameters like fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC). This paper addresses a CEA ejection accident at hot full power (HFP), in which the underlying strong and un-symmetric feedback between thermal-hydraulics and reactor kinetics exist. For this purpose, a multi-physics analysis tool has been selected with the nodal kinetics code, 3DKIN, implicitly coupled to the thermal-hydraulic code, RELAP5, for real-time communication and data exchange. This coupled approach enables high fidelity three-dimensional simulation and is therefore especially relevant to reactivity initiated accident (RIA) scenarios and power distribution anomalies with strong feedback mechanisms and/or un-symmetrical characteristics as in the CEA ejection accident. The Systems Engineering approach is employed to provide guidance in developing the work in a systematic and efficient fashion.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.23-28
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• 1996

The potential for the RELAP5/MOD3.2 was assessed for the loss-of-RHR event during the mid-loop operation and the predictability of major thermal-hydraulic phenomena was also evaluated for the long term transient. The analysis results of the typical two cases(cold leg opening case and pressurizer opening case) were compared with experimental data which was conducted at ROSA-IV/LSTF in Japan. As a result, it was shown that tile code was capable of simulating the thermal-hydraulic transport process with appropriate time step during the reduced inventory operation with the loss-of- RHR system.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.975-981
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• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.647-652
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• 1998

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identified the possible even scenarios following the loss of shutdown cooling. The Thermal hydraulic analyses were performed for the five cases of RCS configurations under the worst event scenario, unavailable secondary cooling and no RCS inventory makeup, using the RELAP5/MOD3.2 code to investigate the plant behavior, From the analyses results, times to boil, times to core uncovery and times to core heat up were estimated to determined the containment closure time to prevent the uncontrolled released of fission products to atmosphere, These data provide useful information to the abnormal procedure to cope with event.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.635-640
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• 1998

Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of Thermal stratification is investigated using Theofanous's empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3787-3800
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• 2023

Reactivity Initiated Accident (RIA) scenarios require special attention using advanced simulation techniques due to their complexity and importance for nuclear power plant (NPP) safety. While the conservative approach has traditionally been used for safety analysis, it may lead to unrealistic results which calls for the use of best estimate plus uncertainty (BEPU) approach, especially with the current advances in computational power which makes the BEPU analysis feasible. In this work an Uncontrolled Control Element Assembly (CEA) Withdrawal at Full Power accident scenario is analyzed using the BEPU approach by loosely coupling the thermal hydraulics best-estimate system code (RELAP5/SCDAPSIM/MOD3.4) to the statistical analysis software (DAKOTA) using a Python interface. Results from the BEPU analysis indicate that a realistic treatment of the accident scenario yields a larger safety margin and is therefore encouraged for accident analysis as it may enable more economic and flexible operation.

• 에너지공학
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• 제17권4호
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• pp.218-226
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• 2008

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. 개발중인 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 REIAP5가 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 수정된 Semi-implicit 방법을 소개하고 이를 이용해 수행한 예비 계산결과를 수록하였다.

• 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.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.977-986
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• 2019

In the thermal-hydraulic system codes, such as MARS and RELAP5/MOD3, the Savannah River Laboratory (SRL) model has been adopted as a subcooled boiling model. It, however, has been shown that the SRL model cannot take into account appropriately the effects of inlet liquid velocity and hydraulic diameter on axial void fraction development. To overcome the problems, Ha et al. (2018) proposed a modified SRL model, which is applicable to low-pressure and low-Pe conditions (P < 9.83 bar and $Pe{\leq}70,000$) only. In this work, the authors extended the modified SRL model by proposing a new net vapor generation (NVG) model and a wall evaporation model so that the new subcooled boiling model can cover a wide range of thermal-hydraulic conditions with pressures ranging from 1.1 to 69 bar, heat fluxes of $97-1186kW/m^2$, Pe of 3600 to 329,000, and hydraulic diameters of 5-25.5 mm. The new model was implemented in the MARS code and has been assessed using various subcooled boiling experimental data. The results of the new model showed better agreements with measured void fraction data, especially at low-pressure conditions.