• 한국유체기계학회 논문집
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• 제18권1호
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• pp.20-28
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• 2015

The pump-turbine impeller is the key component of pumped storage power plant. Current design methods of pump-turbine impeller are private and protected from public viewing. Generally, the design proceeds in two steps: the initial hydraulic design and optimization design to achieve a balanced performance between pump mode and turbine mode. In this study, the 3D inverse design method is used for the initial hydraulic impeller design. However, due to the special demand of high performance in both pump and reverse mode, the design method is insufficient. This study is carried out by modifying the geometrical parameters of the blade which have great influence and need special consideration in obtaining the high performance on the both modes, such as blade shape type at low pressure side (inlet of pump mode, outlet of turbine mode) and the blade lean at blade high pressure side (outlet of pump mode, inlet of turbine mode). The influence of the geometrical parameters on the performance characteristic is evaluated by CFD analysis which presents the efficiency and internal flow results. After these investigations of the geometrical parameters, the criteria of designing pump-turbine impeller blade low and high sides shape is achieved.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1100-1108
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• 2017

A thermal-hydraulic system code is an essential tool for the design and safety analysis of a nuclear power plant, and its accuracy quantification is very important for the code assessment and applications. The fast Fourier transform-based method (FFTBM) by signal mirroring (FFTBM-SM) has been used to quantify the accuracy of a system code by using a comparison of the experimental data and the calculated results. The method is an improved version of the FFTBM, and it is known that the FFTBM-SM judges the code accuracy in a more consistent and unbiased way. However, in some applications, unrealistic results have been obtained. In this study, it was found that accuracy quantification by FFTBM-SM is dependent on the frequency spectrum of the fast Fourier transform of experimental and error signals. The primary objective of this study is to reduce the frequency dependency of FFTBM-SM evaluation. For this, it was proposed to reduce the cut off frequency, which was introduced to cut off spurious contributions, in FFTBM-SM. A method to determine an appropriate cut off frequency was also proposed. The FFTBM-SM with the modified cut off frequency showed a significant improvement of the accuracy quantification.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.60-72
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• 2019

The air-cooled passive decay heat removal system (APDHR) was proposed to provide the ultimate heat sink for non-LOCA accidents. The APDHR is a modified one of Passive Auxiliary Feed-water system (PAFS) installed in APR+. The PAFS has a heat exchanger in the Passive Condensate Cooling Tank (PCCT) and can remove decay heat for 8 h. After that, the heat transfer rate through the PAFS drastically decreases because the heat transfer condition changes from water to air. The APDHR with a vertical heat exchanger in PCCT will be able to remove the decay heat by air if it has sufficient natural convection in PCCT. We conducted the thermal-hydraulic simulation by the MARS code to investigate the behavior of the APR + selected as a reference plant for the simulation. The simulation contains two phases based on water depletion: the early phase and the late phase. In the early phase, the volume of water in PCCT was determined to avoid the water depletion in three days after shutdown. In the late phase, when the number of the HXs is greater than 4089 per PCCT, the MARS simulation confirmed the long-term cooling by air is possible under extended Station Blackout (SBO).

• 시스템엔지니어링학술지
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• 제16권2호
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• pp.67-77
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• 2020

Traditionally nuclear thermal hydraulic and nuclear safety has relied on numerical simulations to predict the system response of a nuclear power plant either under normal operation or accident condition. However, this approach may sometimes be rather time consuming particularly for design and optimization problems. To expedite the decision-making process data-driven models can be used to deduce the statistical relationships between inputs and outputs rather than solving physics-based models. Compared to the traditional approach, data driven models can provide a fast and cost-effective framework to predict the behavior of highly complex and non-linear systems where otherwise great computational efforts would be required. The objective of this work is to develop an AI algorithm to predict the peak fuel cladding temperature as a metric for the successful implementation of FLEX strategies under extended station black out. To achieve this, the model requires to be conditioned using pre-existing database created using the thermal-hydraulic analysis code, MARS-KS. In the development stage, the model hyper-parameters are tuned and optimized using the talos tool.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.131-143
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• 2023

Constitutive equations in a nuclear reactor safety analysis code are mostly empirical correlations developed from experiments, which always accompany uncertainties. The accuracy of the code can be improved by modifying the constitutive equations fitting wider range of data with less uncertainty. Thus, the sensitivity of the code with respect to the constitutive equations is evaluated quantitatively in the paper to understand the room for improvement of the code. A new methodology is proposed which first starts by dividing the thermal hydraulic conditions into multiple sub-regimes using self-organizing map (SOM) clustering method. The sensitivity analysis is then conducted by multiplying an arbitrary set of coefficients to the constitutive equations for each sub-divided thermal-hydraulic regime with SOM to observe how the code accuracy varies. The randomly chosen multiplier coefficient represents the uncertainty of the constitutive equations. Furthermore, the set with the smallest error with the selected experimental data can be obtained and can provide insight which direction should the constitutive equations be modified to improve the code accuracy. The newly proposed method is applied to a steady-state experiment and a transient experiment to illustrate how the method can provide insight to the code developer.

• 한국해안해양공학회지
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• 제18권4호
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• pp.329-337
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• 2006

삼천포 화력발전소에서 냉각수로 이용되고 방류되는 해수를 이용한 소수력 발전소가 삼천포 해역에 건설되고 있다. 본 연구에서는 소수력 발전소가 건설되는 지점, 즉 방류수로 및 방류해역의 흐름을 관측하고 흐름특성을 분석하였다. 방류수로의 흐름은 냉각수 방류량에 의한 영향이 지배적이며, 공간적으로는 냉각수가 합류되는 Weir 상류구간, 합류된 냉각수가 방류수로를 통하여 배출되는 Weir 하류구간, 방류해역으로 구분할 수 있다. Weir 상류 지점은 측면에서 유입되는 냉각수로 인하여 유량이 점차 증가하나, 수로의 폭이 증가하기 때문에 수위 변화가 미미한 지역이며, Weir 하류지점은 월류된 냉각수가 사류(super-critical flow)구간을 거쳐서 수리학적 도약으로 인한 강한 연직방향의 요동을 보이며 방류해역으로 이동하는 구간이다. 한편, 방류수에 의한 영향으로 외해방향으로의 해수흐름이 형성되나 유속은 조위의 승강운동에 크게 영향을 받는 것으로 파악되었다. 또한, 방류해역의 조위는 통영 검조소에 비하여 평균조차는 약 10% 크게 나타나고 있으며, 파랑전파에 의한 영향은 미미한 정도로 파악되었다.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.12-24
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• 2001

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.

• 한국수자원학회:학술대회논문집
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• 한국수자원학회 2015년도 학술발표회
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• pp.296-296
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• 2015

The aim of this paper is to numerically explore the feasibility of designing a Mini-Hydro turbine. The interest for this kind of horizontal axis turbine relies on its versatility. For instance, in the field of renewable energy, this kind of turbine may be considered for different applications, such as: tidal power, run-of-the-river hydroelectricity, wave energy conversion. It is fundamental to improve the turbine performance and to decrease the equipment costs for achievement of "environmental friendly" solutions and maximization of the "cost-advantage". In the present work, the commercial CFD code ANSYS is used to perform 3D simulations, solving the incompressible Unsteady Reynolds-Averaged Navier-Stokes (U-RANS) equations discretized by means of a finite volume approach. The implicit segregated version of the solver is employed. The pressure-velocity coupling is achieved by means of the SIMPLE algorithm. The convective terms are discretized using a second order accurate upwind scheme, and pressure and viscous terms are discretized by a second-order-accurate centered scheme. A second order implicit time formulation is also used. Turbulence closure is provided by the realizable k - turbulence model. In this study, a mini hydro turbine (3kW) has been considered for utilization of horizontal axis impeller. The turbine performance and flow behavior have been evaluated by means of numerical simulations. Moreover, the performance of the impeller varied in the pressure distribution, torque, rotational speed and power generated by the different number of blades and angles. The model has been validated, comparing numerical results with available experimental data.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.702-708
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• 2019

This paper deals with break size estimation of loss of coolant accidents (LOCA) using a nonlinear autoregressive with exogenous inputs (NARX) neural network. Previous studies used static approaches, requiring time-integrated parameters and independent firing algorithms. NARX neural network is able to directly deal with time-dependent signals for dynamic estimation of break sizes in real-time. The case studied is a LOCA in the primary system of Bushehr nuclear power plant (NPP). In this study, number of hidden layers, neurons, feedbacks, inputs, and training duration of transients are selected by performing parametric studies to determine the network architecture with minimum error. The developed NARX neural network is trained by error back propagation algorithm with different break sizes, covering 5% -100% of main coolant pipeline area. This database of LOCA scenarios is developed using RELAP5 thermal-hydraulic code. The results are satisfactory and indicate feasibility of implementing NARX neural network for break size estimation in NPPs. It is able to find a general solution for break size estimation problem in real-time, using a limited number of training data sets. This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr NPP.