• Nuclear Engineering and Technology
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• 제31권1호
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• pp.17-28
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• 1999

A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

• 에너지공학
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• 제21권2호
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• pp.144-151
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• 2012

지역난방시스템이란 열에너지를 생산하고 만들어진 열에너지를 넓은 지역에 공급하는 시스템이다. 지역난방시스템은 국부난방보다 더 효율적이고 친환경적이라고 할 수 있다. 생산 된 열은 단열 배관 네트워크를 통해 사용자들에게 공급된다. 지역난방시스템의 효율적인 운전을 위해서는 여러 가지 운전 조건에서 배관망 네트워크의 유량, 압력 및 온도 분포를 예측하는 것이 중요하다. 본 연구에서는 지역난방용 지중매설 배관망 네트워크에서의 열-유체 동적거동을 수학적 모델을 사용하여 예측하였다. 수학적 모델은 물질, 운동량 및 에너지 수지식을 사용하였다. 모델링의 결과를 검증하기 위하여 강남지역 지역난방시스템에서 측정한 결과와 모델링의 결과를 비교하였다.

• Water Engineering Research
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• 제6권1호
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• pp.1-15
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• 2005

The numerical model, FLUENT, was employed to investigate the effect of the heated water discharged from the diffuser of Boryung Power Plant. Temperature patterns of the thermal effluent discharged from two proposed types of the diffusers was evaluated for maximum flood and maximum ebb tide. The hydraulic model experiments were also performed in the reduced scale of 1/150 to verify the numerical simulation results. The buoyant jets discharged from the diffusers were found to be significantly affected by the ambient flows beyond the region where the effluent momentum was dissipated. Both the numerical and experimental results showed that the area of the excess isotherm for Type 1 diffuser was larger than that for Type 2 diffuser. Type 2 diffuser system was observed to be a more effective diffuser design than Type 1 diffuser system based on the temperature reduction and excess isotherm obtained from the numerical simulation in the ambient flows.

• Corrosion Science and Technology
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• 제11권6호
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• pp.218-224
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• 2012

Liquid droplet impingement erosion (LDIE) known to be generated in aircraft and turbine blades is recently appeared in nuclear piping. UT thickness measurements with both A-scan and B-scan UT inspection equipments were performed for a component estimated as susceptible to LDIE in feedwater heater vent system. The thickness data measured with B-Scan equipment were compared with those of A-Scan. Thermal hydraulic analysis based on ANSYS FLUENT code was performed to analyze the behavior of liquid droplets inside piping. The wall thinning rate and residual lifetime based on both existing Sanchez-Caldera equation and measuring data were also calculated to identify the applicability of the existing equation to the LDIE management of nuclear piping. Because Sanchez-Caldera equation do not consider the feature of magnetite formed inside piping, droplet size, colliding frequency, the development of new evaluation method urgently needs to manage the pipe wall thinning caused by LDIE.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 2003년도 춘계학술대회 논문집
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• pp.105-110
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• 2003

The hydraulic excavator has been a popular research object for automation because of its multi-workings and economic efficiency. When it works crane tasks, most of disasters happen. The objective of this paper is to design each components and to construct boom, arm, bucket circuit. These models modeled with AMESim show us change of variables and behavior of excavator. Simulation model will be used for simulator of excavator.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1540-1555
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• 지질공학
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• 제11권3호
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• pp.255-267
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• 2001

지하 50m의 화강암반내 단층지역에 위치한 지하 방사성폐기물 처분장 인접지역에서의 열, 수리, 및 역학적 연성거동을 비교하고 분석하였다. 해석에는 2차원 해석코드인 UDEC을 사용하였다. 해석모델은 화강암반, 처분공내의 압축 벤토나이트로 둘러싸인 PWR 사용후 핵연료 처분용기, 및 처분동굴내에 채워진 혼합 벤토나이트를 포함한다. 수리-역학적, 열-역학적, 및 열-수리-역학적 연성거동을 비교 및 분석하였다. PWR 사용후 핵연료내의 방사성 물질로부터 나오는 시간의존 방사성 붕괴열이 처분장 및 인접지역에 미치는 영향을 분석하였다. 수리해석에는 steady state flow 알고리즘을 사용하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.177-177
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• 2001