• 에너지공학
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• 제25권1호
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• pp.56-68
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• 2016

최근 원자력 발전소의 안전성을 획기적으로 향상시키기 위한 연구가 활발하게 진행되고 있으며 특히 피동냉각계통의 연구개발이 아주 중요하게 부각되고 있다. 피동냉각계통의 열전달 방식으로는 응축열전달 양식이 주로 채택되고 있다. 이와 같은 맥락에서 부산대학교 Ahn & Yun (Ahn 등, 2014)은 새로운 수평관내부 응축 모델을 제시한 바 있다. 본 연구에서는 먼저 Ahn & Yun 이 제시한 수평관 응축 모델을 MARS 코드에 삽입하고 PASCAL 실험데이터를 이용하여 평가하였다. 이 평가결과를 통해 Ahn & Yun 모델의 코드적용에 있어 문제점을 규명하고 새로운 적용방법론을 적용하여 다양한 실험데이터로 다시 평가함으로써 MARS 코드의 향상된 응축 열전달 해석 능력을 확인하였다.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.759-766
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• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.547-552
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• 1998

The standard RELAP5/MOD3.2 code were assessed with the condensation experiment in the presence of noncondensable gas in a vortical tube of PCCS of CP-1300. There are two wall film condensation models, the default model and the alternative model, in RELAP5/MOD3.2. The experimental apparatus was modeled with the two models, md simulations were performed for several sub-tests to be compared with the experimental results. In overall sense the simulation results showed that the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, while the alternative model over-predicts them throughout the condensing tube.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.559-564
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• 1998

The film condensation models in RELAP5/MOD3.1 and RELAP5/WOD3.2 are assessed with the data experiment performed in the scaled down condensation experimental facility with a single vertical tube inner diameter 46 mm in the range of pressure 0.1∼7.5 Mpa for the PSCS(Passive Secondary Condenser System) Both MOD3.1 and MOD3.2 don't shows any reliable predictions the experimental data The RELAP5/MOD3.1 overpredicts the heat transfer coefficients experiment, whereas the RELAP5/MOD3.2 underpredicts those data it is recommended that the film condonation model in RELAP5/MOD3.2 should be modified to hue a larger heat transfer coefficient than those the present model to give the reliable predictions.

• Nuclear Engineering and Technology
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• 제31권5호
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• pp.465-475
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• 1999

The objective of the present work is to assess the analysis capability of two wall film condensation models, the default and the alternative models, of RELAP5/MOD3.2 on condensation experiments in the presence of noncondensable gas in a vertical tube of PCCS of CP-1300. In the calculation of a base case the default model of RELAP5/MOD3.2 under-predicts the heat transfer coefficients, and Its alternative model over-predicts them throughout the condensing tube, Also, both models over-predict the void fractions. The nodalization study shows that the variation of the node number does not change both modeling results of RELAP5/MOD3.2 Sensitivity study for varying input parameters shows that the inlet steam-air mixture flow rate, the inlet air mass fraction, and the inlet saturated steam temperature give significant changes of their heat transfer coefficients Run statistics show that the grind time of the default model is always higher than that of the alternative model by about 23%.

• 에너지공학
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• 제25권2호
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• pp.1-20
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• 2016

원전의 설계기준사고 및 중대사고 해석에서 응축열전달 모델은 매우 중요하며, 특히 피동냉각계통의 개발이 활발히 진행됨에 따라 그 중요성이 더욱 부각되었다. 그런데, 원자로건물 내부에서와 같이 비응축성기체가 존재하는 경우 응축열전달은 현저히 감소하므로 원전 안전해석에서 이를 고려한 응축열전달 모델이 주목받고 있다. 본 연구에서는 냉각재상실사고 등이 발생하는 경우 원자로건물 내부의 상황과 유사한 열수력 조건에서 수행된 응축열전달 실험자료를 이용하여 중대사고 해석코드 MELCOR 1.8.6의 응축열전달 모델을 평가하였다. 실험조건을 응축면의 형상에 따라 네 가지(수직평판, 수직관 외벽, 수직관 내벽, 수평관 내벽)로 분류하였고, 각 분류별 실험들을 MELCOR 코드로 해석하였다. 해석결과, 수직관 내벽을 제외한 나머지 조건에서 MELCOR 코드가 응축열전달을 전체적으로 저 예측하여 개선이 필요한 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.59-68
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• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.585-590
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• 1997

The condonation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condonation phenomena The default model the laminar film condonation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condonation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP5/MOD3.2 The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to conifer the effects of the gas velocity and the film thickness.

• 한국안전학회지
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• 제36권2호
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.