• 설비공학논문집
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• 제11권1호
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• pp.18-30
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• 1999

A heat transfer analysis for the two-dimensional (2-D) steady state using finite difference method (FDM) is performed to predict the thermal behavior of the primary first-wall (FW) system of fusion reactor under various geometric and thermo-hydraulic conditions, such as the beryllium (Be) armor thickness, pitch of cooling tube, and coolant velocity. The FW consists of authentic steel (type 316 stainless steel solution annealed) for cooling tubes, Cu for cooling tubes embedding material, and Be for a protective armor, based on the International Thermonuclear Experiment Reactor (ITER) report. The present 2-D analysis, the control volume discretized with hybrid grid (rectangular grid and polar grid) and Gauss-Seidel iteration method are adapted to solve the governing equations. In the present study, geometric and thermo-hydraulic parameters are optimized with consideration of several limitations. Consequently, it is suggested that the adequate pitch of cooling tube is 22-32mm, the beryllium armor thickness is 10-12mm, and that the coolant velocity is 4.5m/s-6m/s for $100^{\circ}C$ of inlet coolant temperature. The cooling tube should locate near beryllium armor. But, it would be better for locating the center of Cu wall, considering problems of material and manufacturing. Also, 2-D analysis neglecting the axial temperature distribution of cooling tube is appropriate, regarding the discretization error in axial direction.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.314-327
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• 1999

A subchannel analysis code MATRA applicable to PWRs and ALWRs has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-Rf-1. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and How distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. HATRA has been provided with an improved structure, various functions, and models to give more convenient user environment and to enhance the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the models for the lateral transport between adjacent subchannels have been improved to enhance the accuracy in predicting two-phase flow phenomena. The predictions of MATRA were compared with the experimental data on the flow and enthalpy distribution in some sample rod-bundle cases to evaluate the performance of MATRA. All the results revealed that the predictions of MATRA were better than those of COBRA-IV-I.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.271-276
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• 1996

Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in LMR core steady state thermal-hydraulic performance analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each pin bundle, thus pin cladding damage accrual and pin reliability. The flow orificing analysis for conceptual design will be performed with Excel spreadsheet program ORFCE which was set up and tested, using the calibration factors based on available analyses data. For the verification of this program, flow orificing calculation for the MDP 840MWth core was performed. The calculational results are satisfactory compared to those of CRIEPI calculation.

• 대한기계학회논문집
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• 제9권5호
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• pp.638-646
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• 1985

본 논문에서는 펌프가 동시에 전원차단 되었을 때 송출밸브의 운전조건(개도, 폐쇄속도), 계통의 기하학적 형상(취수 및 배수관로의 길이, 단면적, 재질, 콘덴서의 높이등)과 해면상태(간막의 차에 의한 수위, 파도)의 변화가 계통내 수격현상에 미치 는 영향에 대하여 수치해석 하였다.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1109-1118
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• 2021

The construction of STELLA-2 facility is on-going to demonstrate the safety system of PGSFR and to provide comprehensive understanding of transient behavior under DBEs. Considering that most events are single-phase natural circulation flow with slow transient, STELLA-2 was designed with reduced-height of 1/5 length scale. The ratio of volume to surface area in the vessel can relatively increase resulting in excessive heat transfer. Therefore, a steady-state thermal-hydraulic analysis was performed and the effect of design change to reduce the heat transfer through redan was investigated. The heat transfer through single wall redan in STELLA-2 was 3% of the core power, comparable to 1% of the core power in PGSFR. By applying the insulated redan, about 70% of decrease effect was observed. The effect on transient behavior was also evaluated. The conclusion of this study was directly applied to the STELLA-2 design and the modified version is under construction.

• 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.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• 대한토목학회논문집
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• 제11권4호
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• pp.103-112
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• 1991

본(本) 연구(硏究)는 일정균등(一定均等)한 열적(熱的) 특성(特性)을 가지고 있으며 상변화(相變化) 없는 등방(等方) 이질성(異質性)의 3차원(次元) 대수층계(帶水層系)의 열(熱)흐름과 정상상태(定常狀態)의 지하수(地下水)흐름을 모의발생(模擬發生)할 수 있는 유한차분(有限差分) 모형(模型)을 확립(確立)한 것이다. 이 모형(模型)은 대규모(大規模) 지하수(地下水) 흐름체계(體系)에서 폐기물(廢棄物)의 지하저류시(地下貯溜時) 지하수(地下水) 흐름과 발생(發生) 혹은 주입(注入)된 열(熱)의 흐름을 예측(豫測) 분석(分析)하기 위하여 확립(確立)된 것이다. 이러한 대수층계(帶水層系)의 지하수(地下水) 흐름에 작용(作用)하는 조건(條件)으로는 강우주입(降雨注入)으로 인한 수문학적(水文學的) 조건(條件)과 고정(固定)된 수리수두(水理水頭) 경계조건(境界條件) 등(等)이 포함(包含)되고, 열(熱)흐름에는 지열(地熱)의 흐름, 지표면(地表面)으로의 전도(傳導), 주입(注入)에 의한 이류(移流), 고정(固定) 수두경계(水頭境界)로 향(向)한 또는 고정수두경계(固定水頭境界)로 부터의 이류(移流) 등(等)이 포함(包含)된다. 본(本) 모형(模型)에서는 지하수(地下水)흐름과 열(熱)흐름 방정식(方程式)을 번갈아 푸는 교대반복과정(交代反復過程)을 사용(使用)하고, 두 방정식(方程式)의 계산(計算)에는 직접해법(直接解法)을 사용(使用)한다. 이동시간(移動時間)은 모형공간(模型空間)에서 입자추적(粒子追跡)으로 결정(決定)되며, 분할(分轄)된 구역내(區域內)의 지하수(地下水) 유속(流速)은 구역내(區域內)의 유속(流速)을 선형(線形)으로 보간(補間)하여 계산(計算)한다. 본(本) 모형(模型)을 경상북도(慶尙北道) 영일군(迎日郡) 송라면(松羅面) 지경리(地境里) 일대(一帶)의 지하수계(地下水系)에 적용(適用)하여 이 일대(一帶) 지하(地下) 암반층(岩盤層)의 수두분포(水頭分布), 유동로(流動路), 이동시간(移動時間) 및 지하수온분포(地下水溫分布)를 계산(計算)하여 지하수(地下水) 유동체계(流動體系)를 분석(分析)하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.2141-2145
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• 2004

As part of study on thermal hydraulic behavior in the reactor cavity under external vessel cooling in the APR (Advanced Power Reactor) 1400, one dimensional two phase flow of steady state in the reactor cavity have been analyzed to investigate a coolant circulation mass flow rate in the annulus region between the reactor vessel and the insulation material using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that a two phase natural circulation flow of 300 - 600 kg/s is generated in the annulus region between the reactor vessel and the insulation material when the external vessel cooling has been applied in the APR 1400. An increase in the heat flux of the inner vessel leads to an increase of the coolant mass flow rate. An increase in the coolant outlet area leads to an increase in the coolant circulation mass flow rate, but the coolant inlet area does not effective on the coolant circulation mass flow rate. The change of the lower coolant outlet to a lower position affects the coolant circulation mass flow rate, but the variation trend is not consistent.

• Nuclear Engineering and Technology
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• 제42권6호
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• pp.690-703
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• 2010

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between $5^{th}$ and $95^{th}$ percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure.