• 한국자동차공학회논문집
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• 제3권5호
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• pp.17-23
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• 1995

A numerical analysis on engine coolant is made by the use of FVM based general purpose 3 dimensional Navier-Stokes solver, TURB-3D. Numerical solutions are verified by comparison with the experimental data of Lotus model. The results show a good qualitative as well as quantitative comparison. Coolant flow rate control is attempted through adjusting the cross section area of passage base on the results of an original coolant passage. It is concluded from the results that the flow rate control is possible as attempted, and thus can be used in the real engine design.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1181-1190
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• 2023

Hydraulic performance and flow resistance tests were performed to confirm the main parameters of the hydraulic instrumentation that can affect the pump performance of the reactor coolant pump. The flow resistance test offers important experimental data, which are necessary to predict the behavior of the primary coolant when the circulation of the reactor coolant pump is stopped. Moreover, the shape of the hydraulic section of the pump, which was considered in the test, was prepared to compare the mixed-flow- and axial-flow-type models, the difference in the number of blades of the impeller and diffuser, the difference in the shape of the impeller blade and its thickness, and the effect of coating at the suction bell. Additionally, five models of the hydraulic part were manufactured for the experiments. In this study, the differences in performance owing to the design factors were confirmed through the experimental results.

• 대한기계학회논문집B
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• 제30권8호
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• pp.818-824
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• 2006

The hydrodynamic performance analysis of an axial-flow main coolant pump for the modular nuclear reactor has been carried out using a commercial computational fluid dynamics (CFD) software. The prediction capability of the CFD software adopted in the present study was validated in comparison with the experimental data. Predicted performance curves agree satisfactorily well with the experimental results for the main coolant pump over the normal operating range. π Ie prediction method presented herein can be used effectively as a tool for the hydrodynamic design optimization and assist the understanding of the operational characteristics of general purpose axial-flow pumps.

• 한국유체기계학회 논문집
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• 제2권3호
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• pp.24-29
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• 1999

The ASME strainers have been newly installed at the suction side of each reactor coolant pump to get rid of the foreign materials which may damage the pump impeller or interfere with the coolant path of fuel flow tube or primary plate type heat exchanger. The strainer was designed in accordance with ASME SEC. III, DIV. 1, Class 3 and the structural integrity was verified by seismic analysis. The screen was designed in accordance with the effective void area from the result of flow analysis for T-type strainer. After installation of the strainer, it was confirmed through the field test that the flow characteristics of primary cooling system were not adversely affected. The pressure loss coefficient was calculated by Darcy equation using the pressure difference through each strainer and the flow rate measured during the strainer performance test. And these are useful data to predict flow variations by the pressure difference.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2547-2555
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• 2021

The frequency of reactor coolant leakage is expected to increase over the lifetime of a nuclear power plant owing to degradation mechanisms, such as flow-acceleration corrosion and stress corrosion cracking. When loss of coolant accidents (LOCAs) occur, several parameters change rapidly depending on the size and location of the cracks. In this study, leak flow during LOCAs is predicted using a deep fuzzy neural network (DFNN) model. The DFNN model is based on fuzzy neural network (FNN) modules and has a structure where the FNN modules are sequentially connected. Because the DFNN model is based on the FNN modules, the performance factors are the number of FNN modules and the parameters of the FNN module. These parameters are determined by a least-squares method combined with a genetic algorithm; the number of FNN modules is determined automatically by cross checking a fitness function using the verification dataset output to prevent an overfitting problem. To acquire the data of LOCAs, an optimized power reactor-1000 was simulated using a modular accident analysis program code. The predicted results of the DFNN model are found to be superior to those predicted in previous works. The leak flow prediction results obtained in this study will be useful to check the core integrity in nuclear power plant during LOCAs. This information is also expected to reduce the workload of the operators.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.179-189
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• 1994

이상유동시 원자로 냉각재 펌프의 성능을 펌프의 기하학적 형상 및 단상 유동시의 펌프 성능을 이용하여 예측하였다. 단상 유동시의 원자로 냉각재 펌프의 벽면 마찰손실은 Truckenbrodt의 경계층 이론을 이용하여 예측하였으며, 계산된 벽면 마찰 손실 및 분리 손실을 사용하여 이상유동시의 수두손실을 예측하였다. 해석결과는 Combustion Engineering 사의 펌프 실험 데이터와 비교하였다. 또한 냉각재 상실사고시 이상유동배수가 첨두 피복재 온도에 미치는 영향을 RELAP5를 사용하여 평가하였으며, 분석결과는 이상유동배수의 정확성이 중요한 영향을 미치는 것으로 나타났다.

• 한국산학기술학회논문지
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• 제16권5호
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• pp.3030-3034
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• 2015

새로운 엔진과 차량을 개발하여 엔진을 차량에 탑재할 때, 중요하게 고려해야 할 사항 중의 하나는 냉각 성능이다. 만약 냉각 성능이 열악하다면 엔진은 과열되어 파손되게 된다. 그러나 자동차회사에서 일반적으로 엔진은 차량보다 훨씬 빠른 시기에 개발이 진행되게 되어 엔진을 차량에 탑재한 조건에서 냉각 성능을 시험할 수 없다. 본 연구에서는 몇 가지 시험과 계산 결과를 이용하여 엔진의 냉각 성능을 추정하였다. 첫 번째로 엔진의 열정산 시험이 진행되었다. 두 번째로 냉각수 유동 시험이 진행되었다. 이 시험에서 라디에이터로 유입되는 유량을 구할 수 있다. 그리고 차량의 냉각 시험 성능 조건으로부터 차량의 부하와 속도를 구하고, 이로부터 엔진의 토크와 rpm이 계산되었다. 그리고 이러한 결과를 비교하여 엔진의 냉각 성능이 추정되었다.

• 한국안전학회지
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• 제32권2호
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• pp.147-152
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• 2017

The external reactor vessel cooling (ERVC) is well known strategy to mitigate a severe accident at which nuclear fuel inside the reactor vessel is molten. In order to compare the heat removal capacity of ERVC between the nuclear reactor designs quantitatively, numerical method is often used. However, the study for ERVC using computational fluid dynamics (CFD) is still quite scarce. As a validation study on the numerical prediction for ERVC using CFD, the subcooled boiling flow and natural circulation of coolant at the ULPU-V experiment was simulated. The commercially available CFD software ANSYS-CFX was used. Shear stress transport (SST) model and RPI model were used for turbulence closure and wall-boiling, respectively. The averaged flow velocities in the downcomer and the baffle entry under the reactor vessel lower plenum are in good agreement with the available experimental data and recent computational results. Steam generated from the heated wall condenses rapidly and coolant flows maintains single-phase flow until coolant boils again by flashing process due to the decrease of saturation temperature induced by higher elevation. Hence, the flow rate of coolant natural circulation does not vary significantly with the change of heat flux applied at the reactor vessel, which is also consistent with the previous literatures.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.466-475
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• 2017

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.297-304
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• 2010

In the design of advanced light water reactors (ALWRs) and in the safety assessment of currently operating nuclear power plants, it is necessary to evaluate the possibility of experiencing a degraded core accident and to develop innovative safety technologies in order to assure long-term debris cooling. The objective of this experimental study is to investigate the enhancement factors of dryout heat flux in debris beds by coolant injection from below. The experimental facility consists mainly of an induction heater, a double-wall quartz-tube test section containing a steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of the particle bed was achieved in the radial direction and the axial variation was within 20%. This paper reports the experimental data for 3.2 mm and 4.8 mm particle beds with a 300 mm bed height. The dryout heat density data were obtained for both the top-flooding and the forced coolant injection from below with an injection mass flux of up to $1.5\;kg/m^2s$. The dryout heat density increased as the rate of coolant injection increased. At a coolant injection mass flux of $1.0\;kg/m^2s$, the dryout heat density was ${\sim}6.5\;MW/m^3$ for the 4.8 mm particle bed and ${\sim}5.6\;MW/m^3$ for the 3.2 mm particle bed. The enhancement factors of the dryout heat density were 1.6-1.8.