• Nuclear Engineering and Technology
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• 제18권1호
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• pp.9-16
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• 1986

1984년 11월 14일 원자력 1호기에서 발생된 주급수 상실사고에 대한 계통의 열수력학적인 거동을 모의·해석하고, 발전소 실측자료와의 비교를 통하여 사용된 전산코드의 신뢰도를 평가하였다. 모의된 열수력학적 변수들은 발전소 실측자료와 비교적 잘 일치하였으나 원자로 트립시에 증기발생기 증기유량과 주 냉각재 계통 평균온도에 있어서 약간의 차이를 보였다. 이는 원자로 트립시 깎은 시간에 급격한 노심 출력의 감소로 인하여 열·수력학적 변수들에 큰 변화를 야기하여 발전소 실측자료가 과도상태에서의 불학실성을 내포하기 때문으로 예측되었다. 해석에 사용된 전산코드는 RELAP5/MOD1/CY018로부터 불합리한 oscillation을 일으키는 interphase drag 및 wall heat transfer model의 수정을 통하여 개발된 RELAP5/MOD1/NSC이다.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3164-3182
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• 2023

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

• 드라이브 ㆍ 컨트롤
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• 제16권4호
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• pp.93-100
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• 2019

Nuclear and thermal power plants must provide the turbines with an appropriate degree of high temperature and high pressure steam, to produce the optimum electricity. Additionally, in the event of system and power system failure during electrical production, the steam is immediately disabled, to protect the turbines and generators rotating at high speed. The plant thus uses a special steam control valve system for turbine control, which is opened by force of the hydraulic servo actuator and closed by a large steel spring force. In this study, the causes of failure of the turbine control valve system, a key device of the power plants, were analyzed, and the causes of failure were improved relative to reliability of the equipment.

• Nuclear Engineering and Technology
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• 제47권4호
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• pp.434-442
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• 2015

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

• 한국전산유체공학회지
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• 제22권1호
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• pp.26-36
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• 2017

A numerical study has been performed to evaluate thermal-hydraulic performance of a finger type cooling module with multiple-jet impingement in a divertor of nuclear fusion reactor. To analyze conjugate heat transfer in both solid and fluid domains, numerical analysis of the flow using three-dimensional Reynolds-averaged Navier-Stokes equations has been performed with shear stress transport turbulence model. The computational domain for the cooling module consisted of a single fluid domain and three solid domains; tile, thimble, and cartridge. The numerical results for the temperature variation on the tile were validated in comparison with experimental data under the same conditions. A parametric study was performed with four geometric parameters, i.e., angles between x-axis and centerlines of hole 1, 2, 3 and 4. The results indicate that the heat transfer rate was increased by 2.7% and 0.7% by the angle ${\theta}_1$ and angle ${\theta}_2$, respectively, and that the pressure drop was decreased by up to 1.8% by the angle ${\theta}_3$.

• 한국시뮬레이션학회논문지
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• 제13권2호
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• pp.13-21
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• 2004

The thermal-hydraulic model ARTS which was based on the RETRAN-3D code adopted in the domestic full-scope power plant simulator which was provided in 1998 by KEPRI. Since ARTS is a generalized code to model the components with control volumes, the smaller time-step size should be used even if converged solution could not get in a single volume. Therefore, dedicated models which do not force to reduce the time-step size are sometimes more suitable in terms of a real-time calculation and robustness. In the case of PRT(Pressurizer Relief Tank) model, it is consist of subcooled water in bottom and non-condensable gas in top. The sparger merged under subcooled water enhances condensation. The complicated thermal-hydraulic phenomena such as condensation, phase separation with existence of non-condensable gas makes difficult to simulate. Therefore, the PRT volume can limit the time-step size if we model it with a general control volume. To prevent the time-step size reduction due to convergence failure for simulating this component, we developed a dedicated model for PRT. The dedicated model was expected to provide substantially more accurate predictions in the analysis of the system transients. The results were resonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with the ANSI/ANS-3.5-1998 simulator software performance criteria and RETRAN-3D results.

• 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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• 제53권3호
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• pp.776-792
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• 2021

Conventional integral effect test facilities were constructed to enable the precise observation of thermal-hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal-hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

• 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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• 제54권6호
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• pp.2297-2310
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• 2022

A condensation heat transfer model is essential to accurately predict the performance of the passive containment cooling system (PCCS) during an accident in an advanced light water reactor. However, most of existing models tend to predict condensation heat transfer very well for a specific range of thermal-hydraulic conditions. In this study, a new correlation for condensation heat transfer coefficient (HTC) is presented using machine learning technique. To secure sufficient training data, a large number of pseudo data were produced by using ten existing condensation models. Then, a neural network model was developed, consisting of a fully connected layer and a convolutional neural network (CNN) algorithm, DenseNet. Based on the hold-out cross-validation, the neural network was trained and validated against the pseudo data. Thereafter, it was evaluated using the experimental data, which were not used for training. The machine learning model predicted better results than the existing models. It was also confirmed through a parametric study that the machine learning model presents continuous and physical HTCs for various thermal-hydraulic conditions. By reflecting the effects of individual variables obtained from the parametric analysis, a new correlation was proposed. It yielded better results for almost all experimental conditions than the ten existing models.