• 대한기계학회논문집B
• /
• 제38권3호
• /
• pp.271-277
• /
• 2014

원자로 노심 입구에 위치한 내부 구조물들은 형상 및 노심 입구까지의 상대적 거리에 따라 노심 입구 유량분포에 상당한 영향을 미칠 수 있다. 본 연구에서는 원자로 내부 구조물 형상 처리 방법이 축소 APR+ 유동분포 예측 정확도에 미치는 영향을 조사하기 위해 상용 전산유체역학 소프트웨어인 ANSYS CFX R.14를 사용하여 원자로 내부 구조물들의 실제 형상을 고려한 계산을 수행하였고 다공성 매질 가정을 적용한 계산 결과와 비교하였다. 결론적으로 노심 입구 상류에 위치한 원자로 내부 구조물의 실제 형상을 고려함으로써 노심 입구 유량 분포를 더 정확하게 예측할 수 있었다. 따라서 충분한 계산 자원이 확보된 조건인 경우라면 정확한 노심 입구 유량분포를 계산하기 위해 노심 입구 상류에 위치한 원자로 내부 구조물들(예: 하부지지구조물 바닥판 및 노내 계측기 노즐 지지판)의 실제 형상을 고려해서 계산하는 것이 필요하다.

• Nuclear Engineering and Technology
• /
• 제27권4호
• /
• pp.562-570
• /
• 1995

근래의 ABB/CE형 가압경수로들은, 정상운전 및 예상운전과도상태 중에 허용핵연료설계제한치가 위배되는 것을 방지하기 위하여, 노심 운전상태를 감시하는 디지탈노심감시계통, COLSS(Core Oper-ating Limit Supervisory System)를 보유하고있다. COLSS의 주요 기능 중 하나는. 측정되는 운전조건에 대한 최소 핵비등이탈률을 계산하여, 핵비등이탈에 대한 과출력여유도를 감시하는 것이다. COL-SS에서 최소 핵비등이탈률을 계산하는데 사용되는 CETOP-D 모델은 상세부수로분석코드인 TORC 모델에 대해 보수적으로 벤치마킹되며, 보정상수로서 고온집 합체의 입구유량인자를 사용하고 있다. 본 연구에서는 CETOP-D 입구유량인자를 가장 제한적인 운전조건에서 보수적인 단일 값으로 결정하는 ABB/CE 방법을 배제하고, 운전조건에 따른 CETOP-D 입구유량인자 변화를 상관식형태로 결정하는 “CETOP-D 입구유량인자 최적화기법”을 개발하였다. 개발된 방법을 영광 3,4호기 초기노심의 노심운전영역에 적용한 결과, 기존의 ABB/CE 방법에 비하여 정상운전영역에서 핵비등이탈에 대한 과출력여유도가 2% 가량 증가하였다.

• Nuclear Engineering and Technology
• /
• 제23권3호
• /
• pp.340-351
• /
• 1991

l/5.03 축소 원자로 모델을 이용하여 원자력 발전소 영광 3,4호기를 위한 유동시험을 수행하였다. 이 유동 시험의 목적은 ABB-CE사의 System 80과 영광 3,4호기 원자로 크기의 상대적인 차이로 인해 발생하는 원자로 용기내의 수력학적 영향을 평가하는 것이다. 유동 모델은 상사성 원리에 따라 설계하였다. 이 시험에서 얻은 결과는 노심 입구 유량 분포, 노심 출구 압력 분포, 원자로 입구 노즐에서부터 출구 노즐까지 유동로를 따른 부분 구간 및 전체 압력 손실이다. 이 데이터들은 노심의 열적 여유도 분석에 필요한 입력 자료 제공과 해석적 수력설계 방법의 검증에 이용하게 된다.

• 설비공학논문집
• /
• 제18권10호
• /
• pp.819-825
• /
• 2006

In a SOFC/GT (solid oxide fuel cell/gas turbine) hybrid power generation system, the recuperator is an indispensible component to enhance system performance. Since the expansion ratio to the recuperator core is very large, generally, the effective header design to distribute the flow uniformly before entering the core is crucial to guarantee the required performance. In the present study, we focus on the design of a diffuser type recuperator header with a 90 degree turn inlet port. To reduce the flow separation and recirculation flows, multiple horizontal vanes are used. The number of horizontal vanes is varied from 0 to 24. The air flow velocity is measured at 40 points just behind the core outlet by using a hot wire anemometer. Then, the flow non-uniformity is evaluated from the measured flow velocity. The experimental results showed that inlet air velocity did not effect on relative flow non-uniformity. According to increasing the number of horizontal vanes, flow non-uniformity reduced about $40{\sim}50%$ than without using horizontal vanes.

• Nuclear Engineering and Technology
• /
• 제54권9호
• /
• pp.3567-3579
• /
• 2022

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

• Nuclear Engineering and Technology
• /
• 제48권4호
• /
• pp.941-951
• /
• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• 한국유체기계학회 논문집
• /
• 제13권1호
• /
• pp.35-41
• /
• 2010

Experimental study on the flow field inside the nozzle for radial turbine was performed. At design point, the pressure is high and the Mach number is low at the pressure side of the nozzle inlet semi-vaneless space as the flow turns through the nozzle vanes. As the flow accelerates through the nozzle passage to the throat the pressure level at the pressure and suction sides becomes similar. The flow continued accelerating from the throat to the inlet of turbine wheel and the pressure field became uniform in the circumferential direction in the vaneless space. In high expansion ratio condition, strong favorable pressure gradient band region occurred just after the throat in the semi-vaneless space in the circumferential direction and the pressure became uniform in the circumferential direction after this band. In low expansion ratio condition, core flow acceleration is dominant after the throat and this non-uniform pressure field reached to the inlet of turbine wheel.

• Nuclear Engineering and Technology
• /
• 제54권2호
• /
• pp.617-626
• /
• 2022

To effectively monitor the variety of distributions of neutron flux, fuel power or temperatures in the reactor core, usually the ex-core and in-core neutron detectors are employed. The thermocouples for temperature measurement are installed in the coolant inlet or outlet of the respective fuel assemblies. It is necessary to reconstruct the measurement information of the whole reactor position. However, the reading of different types of detector in the core reflects different aspects of the 3D power distribution. The feasibility of reconstruction the core three-dimension power distribution by using different combinations of in-core, ex-core and thermocouples detectors is analyzed in this paper to synthesize the useful information of various detectors. A comparison of multilayer perceptron (MLP) network and radial basis function (RBF) network is performed. RBF results are more extreme precision but also more sensitivity to detector failure and uncertainty, compare to MLP networks. This is because that localized neural network could offer conservative regression in RBF. Adding random disturbance in training dataset is helpful to reduce the influence of detector failure and uncertainty. Some convolution neural networks seem to be helpful to get more accurate results by use more spatial layout information, though relative researches are still under way.

• Nuclear Engineering and Technology
• /
• 제38권2호
• /
• pp.109-118
• /
• 2006

Various indirect power cycle options for a helium cooled gas cooled fast reactor (GFR) with particular focus on a supercritical $CO_2(SCO_2)$ indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The balance of plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and $SCO_2$ recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of $550^{\circ}C$, (2) advanced design with turbine inlet temperature of $650^{\circ}C$ and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect $SCO_2$ recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR &proximate-containment& and the BOP for the $SCO_2$ cycle is very compact. Both these factors will lead to reduced capital cost.

• Nuclear Engineering and Technology
• /
• 제27권5호
• /
• pp.710-720
• /
• 1995

원자로에서 펌프에 의해 야기되는 맥동 압력은 원자로 내부 구조물에 진동과 손상을 줄 수 있기 때문에 관심이 증가되고 있다. 본 연구에서는 냉각관과 환형관(원자로 압력 용기와 노심 보호 지지대 사이)으로 구성된 기하 형태에서 펌프에 의해 야기되는 맥동 압력을 해석할 수 있는 수력학적 모델을 개발하였다. 수학적 지배 방정식은 압축성, 비점성 유체에 대해 선형화된 Navier-Stokes 방정식이다. 냉각관과 환형관을 따로 분리하여 해석하고 두영역의 커플링 영향을 고려하였다. 또한 본 기하 형태에서 펌프맥동 압력에 영향을 미치는 주요 기하 인자에 대한 평가를 수행하였다. 본 해석 결과와 실험차를 비교하여 만족할 만한 결과를 얻었다.