• 한국연소학회:학술대회논문집
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• 한국연소학회 2014년도 제49회 KOSCO SYMPOSIUM 초록집
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• pp.165-166
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• 2014

The application of particle reaction model in the packed bed process modeling is discussed for iron ore pellet induration process. Combustion of coke breeze in the pellet is estimated by using shrinking unreacted-core model and grain model in which the progress of chemical reaction is described in different concepts. Under the identical inlet gas and solid conditions, the calculation using shrinking core model showed deviated results in terms of temperature profile and conversion fraction, which may imply the significance of selecting proper particle reaction model in consideration of particle characteristics and process operation conditions.

• 한국유체기계학회 논문집
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• 제15권4호
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• pp.5-11
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• 2012

To identify the flow characteristics of the SMART reactor, a flow distribution model test and a numerical simulation are performed in KAERI. Among several part of the SMART reactor, the fuel assemblies are simulated using simulators because of the complexity. The geometries of the core in the SMART reactor and simulator are different, but some similarities are maintained such as the ratio of pressure drop in the vertical and cross directions. There are cross flow holes in each core simulator to reproduce the cross flow of SMART fuel assemblies. To know the flow characteristics of the cross flow, numerical analysis is performed. As the cross flow area is decreased, the pressure drop between inlet and outlet is decreased. Also, when the flow imbalance between two core simulators is constant, the cross flow area does not significantly affect the cross flow.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• 대한기계학회논문집B
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• 제37권9호
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• pp.855-862
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• 2013

개방 노심 열적여유도 해석 코드에 입력으로 제공되는 APR+ (Advanced Power Reactor Plus)의 수력학적 특징을 결정하기 위해 일련의 1/5 축소 원자로 유동분포 시험이 수행되었다. 본 연구에서는 원자로 내부 유동 계산시 다공성 모델을 사용한 전산유체역학의 적용성을 평가하기 위해 상용 전산유체역학 소프트웨어인 ANSYS CFX V.14를 사용하여 계산을 수행하였다. 결론적으로 본 연구에서 사용한 일부 원자로 내부 구조물에 대한 다공성 영역 처리방식을 통해 원자로 내부의 유동 특성을 정성적으로 적절히 파악할 수 있을 것으로 판단된다. 만일 충분한 계산 자원이 확보된 조건인 경우라면 노심 입구 상류에 위치한 원자로 내부 구조물의 실제 기하 형상을 고려함으로써 노심 입구 유량분포를 보다 정확하게 예측할 수 있을 것으로 예상된다.

• 설비공학논문집
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• 제29권12호
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• pp.654-662
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• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3275-3285
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• 2021

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

• Nuclear Engineering and Technology
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• 제14권2호
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• pp.51-63
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• 1982

한국에너지연구소에서 개발한 1차원적 제논과도현상해석 코드 DD1D를 사용하여 가압경수로심의 축방향 제논진동에 대한 안정성을 조사하였다. 노심의 출력준위, 감속재온도계수, 노심 입구온도, 도플러출력 계수 그리고 연소도의 변화가 노심의 축방향 안정성에 미치는 효과를 조사하기 위하여 고리1호기의 설계 및 운전자료를 이용하였으며 본 민감도 분석을 통하여 고리 1호기의 노심은 주기 초에는 축방향 제논진동에 대하여 안정하나 연소도가 증가함에 따라 안정도가 차츰 감소하여 주기 말에는 불안정해진다는 것을 알았다. 이같이 연소도가 증가함에 따라 노심의 안정도가 감소하는 이유는 연소도 변화에 따라 축방향의 출력분포, 감속재온도 계수 및 도플러출력계수가 변하기 때문이다. 본 연구를 통하여 출력밀도가 높은 대형 가압 경수로의 경우 전 주기동안 축방향제논진동에 대하여 안정된 노심을 설계하기 힘들다는 결론에 도달하였다.

• 설비공학논문집
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• 제1권4호
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• pp.265-275
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• 1989

An experimental study of decaying swirl air flow has been obtained by tangential inlet in a straight tube with Reynolds number range 20,000~40,000. The friction factor, swirl angle, velocity profiles and turbulent intensity are measured by using micro-manometer and hot-wire anemometer. It is found that the swirl flow behaviors depend on the swirl intensity along the test tube.

• 설비공학논문집
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• 제11권1호
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• pp.81-90
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• 1999

The performance evaluation technique of a heat exchanger is described by using a transient response analysis for the determination of an average heat transfer coefficient. The model using a finite difference method can accommodate arbitrary inlet fluid temperature as well as longitudinal conduction. Temperature histories are obtained from the experiments at the inlet and outlet of test core. Heat transfer coefficient and friction factor of the plate array are obtained in short times using the data reduction program of transient response analysis in the single-blow method. The results agree very well with theoretical results. It is shown that the rms deviations are very small and the performance evaluation technique gives rapid and accurate results.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1417-1428
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• 2020

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MW_th. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MW_th.