• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3526-3539
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• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to simulate nuclear reactor systems, which solve simplified governing equations by replacing source terms with constitutive relations for simulating entire reactor systems with low computational resources. For half a century, many efforts have been made for wider versatility and higher accuracy of system codes, but various factors can affect the code analysis results, and it was difficult to isolate these factors and interpret them individually. In this study, two system codes, RELAP5 and TRACE, which have many users and are highly reliable, are selected to analyze only the effects of constitutive relations. The influence of constitutive relations is analyzed using in-house platforms that replicate constitute relations of RELAP5 and TRACE equally to exclude factors that may affect analysis results, such as governing equation solvers and user effects. Among the various constitutive relations, the analysis is performed on the wall variables expected to have the most influence on the analysis results. Part 1 paper presents the methodology and wall friction model comparison, while Part 2 paper shows wall heat transfer comparison of the two selected codes.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.131-143
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• 2023

Constitutive equations in a nuclear reactor safety analysis code are mostly empirical correlations developed from experiments, which always accompany uncertainties. The accuracy of the code can be improved by modifying the constitutive equations fitting wider range of data with less uncertainty. Thus, the sensitivity of the code with respect to the constitutive equations is evaluated quantitatively in the paper to understand the room for improvement of the code. A new methodology is proposed which first starts by dividing the thermal hydraulic conditions into multiple sub-regimes using self-organizing map (SOM) clustering method. The sensitivity analysis is then conducted by multiplying an arbitrary set of coefficients to the constitutive equations for each sub-divided thermal-hydraulic regime with SOM to observe how the code accuracy varies. The randomly chosen multiplier coefficient represents the uncertainty of the constitutive equations. Furthermore, the set with the smallest error with the selected experimental data can be obtained and can provide insight which direction should the constitutive equations be modified to improve the code accuracy. The newly proposed method is applied to a steady-state experiment and a transient experiment to illustrate how the method can provide insight to the code developer.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.11-32
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• 2006

System thermal-hydraulic codes have been used in the past decades in the areas of design, operation, licensing and safety of Nuclear Power Plants (NPPs). The development and validation of these codes have reached a high degree of maturity, through the consideration of huge experiments and advanced numerical models. Nowadays, the analyses are based upon realistic approaches rather than the conservative evaluation models. However the applications of these computational tools require preliminary qualification issues. Although huge amounts of financial and human resources have been invested for the development and improvement of codes, the calculation results are still affected by errors. In the sophisticated nuclear technology, design and safety of NPP, these errors must be quantified. An overview of the state of the art of the current thermal-hydraulic system code is developed and the need of uncertainty analysis in code calculations is emphasized. Several sources of uncertainty have been classified and commented, and typical applications of such methods are shown.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.39-42
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• 2002

The nuclear transmutation technology to Incinerate the long lived radioactive nuclides and produce energy during the incineration process is believed to be one or the best solutions. HYPER(${\underline{HY}}brid {\underline{P}}ower {\underline{E}}xtraction {\underline{R}}$eactor)is the accelerator driven transmutation system which is being developed by KAERI(Korea Atomic Energy Research Institute). Lead-bismuth(Pb-Bi) is adopted as a coolant and spallation target material. In this paper, we performed the thermal-hydraulic analysis of HYPER target using the commercial code FLUENT, and also calculated thermal and mechanical stress of the beam window using the commercial code ANSYS. It is found that there is an optimum value for the window diameter and the maximum allowable beam current can be increased to 17.3 mA for the inner diameter of windows, 40 cm. Finally, the other shapes such as uniform or scanned beam were considered. The results of FLUENT calculations show that the uniform type is preferable to the other shapes of the beam in terms of the window and target cooling and the maximum window temperature is lower than that of the parabolic beam by $58 ^{\circ}C$ for the beam current, 13 mA.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.225-236
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• 1994

SEBIM 밸브 상부에 위치한 밀봉수의 급격한 방출은 밸브 후단의 방출배관계통에 큰 운동량과 관성력의 작용을 초래한다. 본 연구는 밸브개방시 방출배관계통의 후단에 발생하는 열수력학적 과도현상을 분석하기 위한 해석절차 및 해석결과를 다루고 있으며, 이 분석을 위해 RELAP5 /MOD3 를 사용하였다. RELAP5 /MOD3 분석을 위하여, 방출관 계통과 SEBIM 밸브의 개방특성 및 밀봉수 방출등의 적절한 모델방법이 제시되었다. 또한 접합부(junction)와 체적(volume)의 제어 플래그 (flag)에서 옵션(option)의 적절한 선택을 위하여 민감도분석도 수행되었다. 분석결과, SEBIM 밸브 방출배관계통의 밀봉수 방출에 따른 열수력학적 과도현상을 분석하는데 RELAP5 /MOD3가 적절히 사용될 수 있음을 알 수 있었다. 민감도 분석결과로부터, 밀봉수 방출해석을 위해서는 적절한 기하학적 압력분포를 가지는 완만한(smooth) 면적변화 및 비평형 옵션(option), 적절한 시간간격(time step)의 사용이 필수적인 것을 알 수 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.427-428
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• 2004

• 유공압시스템학회논문집
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• 제4권2호
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• pp.28-33
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• 2007

In the power plant facility which use soft coal as a power source the fan pitch blade control hydraulic actuator is used to control the inlet and outlet gas to regulate the internal pressure of the furnace and control the frequence. Sometimes malfunctions of this equipment lead to the decline of boiler thermal efficiency and unexpected power plant trip. In order to localize the fan pitch blade control hydraulic actuator specially for the 500MW large scale boiler, Analysis and modelling of the system is carried out mathematically. The responses of the system are examined by using matlab simulation fur the variation of the major parameters in view of reverse engineering. Consequently the validity of the established parameters are examined.

• 한국지반공학회논문집
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• 제29권8호
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• pp.65-73
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• 2013

지중 열교환 시스템은 지속적인 에너지 효율의 개선으로 공간 냉난방을 위한 친환경적 에너지 기술로 주목받고 있다. 지중에 매설된 파이프는 내부 유체 순환을 통하여 인접한 지반과 열적 상호작용으로부터 직접적인 열에너지 교환을 수행한다. 하지만, 파이프의 수치모델링에서 열-수리가 연관된 난류해석과 파이프의 긴 세장비에 의한 메쉬사이즈의 부적합성은 열교환 시스템의 적절한 수치해석을 어렵게 하고 있다. 본 논문에서는 파이프 내부 유체흐름에 대한 에너지 보존의 법칙을 적용하여 지배방정식을 유도하였으며, Galerkin수식화와 시간적분을 통하여 열-수리 연동일차원 파이프 요소를 개발하였다. 그리고 제안된 파이프 요소를 기 개발된 다공질 재료를 위한 열-수리-역학(Thermo-Hydro-Mechanical) 해석을 위한 유한요소 프로그램과 결합하였다. 개발된 요소를 이용한 수치해석 결과는 열응답 시험(Thermal Response Test) 결과로부터 주위지반의 유효 열전도도를 평가하기 위하여 사용하는 선형 열원 모델이 인접 파이프간의 열적상호작용과 파이프의 단부효과에 의하여 지반의 열전도도를 과다 평가하는 것으로 보여주었다. 따라서 열응답 시험 해석 결과에 대한 역해석을 적용하여 최적의 수렴성을 보여주는 변환행렬을 제시하였다.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.