• 한국전산유체공학회지
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• 제21권3호
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• pp.89-97
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• 2016

In this study, direct code coupling, in which two codes share a single flow field, was conducted using 3-dimensional high resolution thermal hydraulics code, CUPID and 1-dimensional system analysis code, MARS. This approach provide the merit to use versatile capability of MARS for nuclear power plants and 3-dimensional T/H analysis capability of CUPID. Numerical Method to directly couple CUPID and MARS was described in this paper. The straight flow and manometer flow oscillation were calculated to verify conservation of coupled CUPID/MARS code in mass, momentum, and energy. This verification calculations indicates that the CUPID/MARS is coupled appropriately in numerical aspect and the coupled code can be applied to nuclear reactor thermal hydraulics after validation against integral transient experiments.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.587-594
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• 2005

In an effort to develop a thermal-hydraulic (TH) safety analysis code for Gas-cooled Reactors (GCRs), the MARS code, which was primarily developed for TH analysis of water reactor systems, has been extended here for application to GCRs. The modeling requirements of the system code were derived from a review of major processes and phenomena that are expected to occur during normal and accident conditions of GCRs. Models fur code improvement were then identified through a review of existing MARS code capability. Among these, the following priority models necessary fur the analysis of limiting high and low pressure conduction cooling events were evaluated and incorporated in MARS-GCR/V1 : 1) Helium (He) and Carbon Dioxide ($CO_2$) as main system fluids, 2) gas convection heat transfer, 3) radiation heat transfer, and 4) contact heat transfer models. Each model has been assessed using various conceptual problems for code-to-code benchmarks and it was demonstrated that MARS-GCR/V1 is capable of capturing the relevant phenomena. This paper describes the models implemented in MARS-GCR/V1 and their verification and validation results.

• 에너지공학
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• 제14권4호
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• pp.232-240
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• 2005

경수 및 중수로 원자로계통에 대한 열수력 안전해석을 위해 개발된 MARS 코드가 고온가스로에 적용될 수 있을지 화인하기 위하여 IAEA TECDOC-1163에서 제시된 고온가스로 원자로공동냉각계통에 대한 Benchmark problem을 평가계산 하였다. HTR-10과 HTTR의 MARS 코드 계산결과는 기 보고된 THERMIX 코드와 THANPACST2 코드의 계산결과 그리고 가용한 실험결과와 비교한 바, 최대 오차범위 $4.5\%$ 정도로 전반적으로 일치하는 것으로 나타났다. 오차의 주요 원인은 복잡한 기하학적 구조를 단순하게 모델링한 부분과 MARS 코드에서 모사하기 어려운 냉각기 , 공기냉각기와 같은 고온가스로. Component에서 발생하였다. 경수형 원자로에서는 중요하게 고려하지 않았던 복사열전달이 고온가스로 원자로공동에서는 붕괴열 제거에 중요한 역할을 수행하는 것으로 나타났다. 결론적으로, 본 연구를 종합하여 볼 때 MARS 코드는 고온가스로 원자로공동냉각계통의 냉각능력을 잘 모사하고 있으며 향후 수소생산용 고온가스로 개발에 있어서 안전해석 코드로서의 역할을 충분히 수행할 수 있을 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.344-363
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• 1999

A multi-dimensional thermal-hydraulic system analysis code, MARS 1.3.1, has been developed in order to have the realistic analysis capability of two-phase thermal-hydraulic transients for pressurized water reactor (PWR) plants. As the backbones for the MARS code, the RELAP5/MOD3.2.1.2 and COBRA-TF codes were adopted in order to take advantages of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the MARS code, all the functional modules of the two codes were unified into a single code first. Then, the source codes were converted into the standard Fortran 90, and then they were restructured using a modular data structure based on "derived type variables" and a new "dynamic memory allocation" scheme. In addition, the Windows features were implemented to improve user friendliness. This paper presents the developmental work of the MARS version 1.3.1 including the hydrodynamic model unification, the heat structure coupling, the code restructuring and modernization, and their verifications.their verifications.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1015-1024
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• 2009

The presence of a non-condensable gas can considerably reduce the level of condensation heat transfer. The non-condensable gas effect is a primary concern in some passive systems used in advanced design concepts, such as the Passive Residual Heat Removal System (PRHRS) of the System-integrated Modular Advanced ReacTor (SMART) and the Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR). This study examined the capability of the Multi-dimensional Analysis of Reactor Safety (MARS) code to predict condensation heat transfer in a vertical tube containing a non-condensable gas. Five experiments were simulated to evaluate the MARS code. The results of the simulations showed that the MARS code overestimated the condensation heat transfer coefficient compared to the experimental data. In particular, in small-diameter cases, the MARS predictions showed significant differences from the measured data, and the condensation heat transfer coefficient behavior along the tube did not match the experimental data. A new method for calculating condensation heat transfer coefficient was incorporated in MARS that considers the interfacial shear stress as well as flow condition determination criterion. The predictions were improved by using the new condensation model.

• 에너지공학
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• 제23권4호
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• pp.112-122
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• 2014

A main steam line break (MSLB) test at the ATLAS facility was simulated using the best-estimate thermal-hydraulic system code, MARS-KS. This has been performed as an activity at the third domestic standard problem for code benchmark (DSP-03) that has been organized by Korea Atomic Energy Research Institute (KAERI). The results of the MSLB experiment and the MARS input data prepared for the previous DSP-02 using the ATLAS facility were provided to participants. The preliminary MSLB simulation using the base input data, however, showed unphysical results in the primary-to-secondary heat transfer. To resolve the problems, some improvements were implemented in the MARS input modelling. These include the use of fine meshes for the bottom region of the steam generator secondary side and proper thermal-hydraulics calculation options. Other input model improvements in the heat loss and the flow restrictor models were also made and the results were investigated in detail. From the results of simulations, the limitations and further improvement areas of the MARS code were identified.

• Advances in aircraft and spacecraft science
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• 제8권1호
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• pp.1-15
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• 2021

Mars exploration demands aerodynamic computations for a proper design of missions of spacecraft carrying instruments and astronauts to Mars. Both Computational Fluid Dynamics (CFD) and Direct Simulation Monte Carlo (DSMC) method play a key role for this purpose. To the author's knowledge, the altitude separating the fields of applicability of CFD and DSMC in Mars atmosphere entry is not yet clearly defined. The limitations in using DSMC at low altitudes are due to technical limitations of the computer. The limitations in using CFD at high altitudes are due to thermodynamic non-equilibrium. Here, this problem is studied in Mars atmosphere entry, considering the Mars Pathfinder capsule in the altitude interval 40-80 km, by means of a DSMC code. Non-equilibrium is quantified by the relative differences between translational temperature and: rotational (θt-r), vibrational (θt-v), overall (θt-ov) temperatures, anisotropy is quantified by the relative difference between the translational temperature component along x and those along y (θx-y) and along z (θx-z). The results showed that θt-r, θt-v, θx-y, θx-z are almost equivalent. The altitude of 45 km should be the limit altitude for a proper use of a CFD code and the altitude of 40 km should be the limit altitude for a reasonable use of a DSMC code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.157-157
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• 1998

• 한국안전학회지
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• 제27권5호
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• pp.229-234
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• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• 에너지공학
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• 제27권4호
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• pp.27-35
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• 2018

피동원자로건물냉각계통(PCCS)은 사고 발생 시 원자로건물로 방출된 열을 제거하여 원전의 건전성을 보장하기 위해 설계되었다. PCCS의 열제거 성능은 증기-공기 혼합물의 응축열전달에 의해 결정된다. 본 연구에서는 응축열전달계수의 예측 정확도를 향상시키기 위해 새로운 상관식을 이식한 MARS-KS 코드를 사용하여 PCCS의 열제거 성능을 평가하였다. MARS-KS 코드에 사용된 새로운 상관식은 압력, 벽면과냉도, 비응축성 기체 질량분율 및 응축튜브의 종횡비와 같은 열전달계수에 영향을 미치는 변수들을 이용하여 개발하였고, 이는 MARS-KS코드의 기본 응축 모델인 Colburn-Hougen 모델을 대체하여 적용되었다. 대형파단 냉각재상실사고 발생 시 PCCS의 운전에 따른 다양한 열수력학적 변수들을 분석하였고, 열제거 성능 평가를 위해 새로운 상관식이 적용된 MARS-KS 코드의 원자로건물 압력거동 계산결과와 기존의 응축모델을 이용한 해석결과를 비교하였다.