• 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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• 제56권5호
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• pp.1687-1697
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• 2024

This study presents the Supporting platform for reactor fine flow characteristics calculation and analysis (Cilian platform), a user-friendly tool that supports the analysis and optimization of pressurized water reactor (PWR) cores with mixing vanes using computational fluid dynamics (CFD) computing. The Cilian platform allows for easy creation and optimization of PWR's main CFD calculation schemes and autonomously manages CFD calculation and analysis of PWR cores, reducing the need for human and computational resources. The platform's key features enable efficient simulation, rapid solution design, automatic calculation of core scheme options, and streamlined data extraction and processing techniques. The Cilian platform's capability to call external CFD software reduces the development time and cost while improving the accuracy and reliability of the results. In conclusion, the Cilian platform exemplifies an innovative solution for efficient computational fluid dynamics analysis of pressurized water reactor (PWR) cores. It holds great promise for driving advancements in nuclear power technology, enhancing the safety, efficiency, and cost-effectiveness of nuclear reactors. The platform adopts a modular design methodology, enabling the swift and accurate computation and analysis of diverse flow regions within core components. This design approach facilitates the seamless integration of multiple computational modules across various reactor types, providing a high degree of flexibility and reusability.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• 제27권6호
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• pp.845-852
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• 1995

반복계산법 (iterative method)중의 하나인 파형완화법 ( Waveform Relaxation Method)을 이용하여 시간의 함수인 비선형 원자로 동역학(reactor dynamics)의 해를 병렬처리기법으로 구하였다. 파형완화법은 각 반복계산과정중에서, 먼저 전체 시스템을 작은 부시스템(subsystem)으로 나누고, 각각의 부시스템들은 주어진 각각의 시간간격(time Interval)에서 해를 독립적으로 구한다. 만약 이 기법에 맞게 효과적으로 부시스템으로 나눌 수 있다면 병렬처리기법에 잘 응용될 수 있다 본 논문에서는 파형완화법이 소개되었고, 두 종류의 원자로 동역학에 응용되었다. 결론적으로 파형완화법은 원자로 동역학에 응용될 수 있으나, 다목적 연구로에 응응한 결과는 병렬효과가 그다지 크지 않은 것으로 나타났다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.154-154
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• 2001

• 전기의세계
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• 제27권6호
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• pp.58-62
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• 1978

The problem of optimal shutdown control of nuclear reactor having nonlinear dynamics is considered. Since the problem, being a bounded state space problem, is difficult to solve by conventional analytic methods such as Pontryagin's maximum principle, it is approached directly by the quasilinearization technique, and solved numerically. The solution obtained in this manner proves to be an improvement over the previous results.

• 한국지능시스템학회논문지
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• 제5권3호
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• pp.52-63
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• 1995

A neuro-fuzzy algorithm is presented for nuclear reactor power control in a pressurized water reactor. Automatic reacotr power control is complicated by the use of control rods because of highly nonlinear dynamics in the axial power shape. Thus, manual shaped controls are usually employed even for the limited capability during the power maneuvers. In an attempt to achieve automatic shape control, a neuro-fuzzy approach is considered because fuzzy algorithms are good at various aspects of operator's knowledge representation while neural networks are efficinet structures capable of learning from experience and adaptation to a changing nuclear core state. In the proposed neuro-fuzzy control scheme, the rule base is formulated based ona multi-input multi-output system and the dynamic back-propagation is used for learning. The neuro-fuzzy powere control algorithm has been tested using simulation fesponses of a Korean standard pressurized water reactor. The results illustrate that the proposed control algorithm would be a parctical strategy for automatic nuclear reactor power control.

• Nuclear Engineering and Technology
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• 제39권3호
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• pp.221-230
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• 2007

Wavelet theory was applied to detect a singularity in a reactor power signal. Compared to Fourier transform, wavelet transform has localization properties in space and frequency. Therefore, using wavelet transform after de-noising, singular points can easily be found. To test this theory, reactor power signals were generated using the HANARO(a Korean multi-purpose research reactor) dynamics model consisting of 39 nonlinear differential equations contaminated with Gaussian noise. Wavelet transform decomposition and de-noising procedures were applied to these signals. It was possible to detect singular events such as a sudden reactivity change and abrupt intrinsic property changes. Thus, this method could be profitably utilized in a real-time system for automatic event recognition(e.g., reactor condition monitoring).

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1776-1783
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• 2019

To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

• Nuclear Engineering and Technology
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• 제50권5호
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• pp.665-672
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• 2018

A venturi scrubber is an important element of Filtered Containment Venting System (FCVS) for the removal of aerosols in contaminated air. The present work involves computational fluid dynamics (CFD) study of dust particle removal efficiency of a venturi scrubber operating in self-priming mode using ANSYS CFX. Titanium oxide ($TiO_2$) particles having sizes of 1 micron have been taken as dust particles. CFD methodology to simulate the venturi scrubber has been first developed. The cascade atomization and breakup (CAB) model has been used to predict deformation of water droplets, whereas the Eulerian-Lagrangian approach has been used to handle multiphase flow involving air, dust, and water. The developed methodology has been applied to simulate venturi scrubber geometry taken from the literature. Dust particle removal efficiency has been calculated for forced feed operation of venturi scrubber and found to be in good agreement with the results available in the literature. In the second part, venturi scrubber along with a tank has been modeled in CFX, and transient simulations have been performed to study self-priming phenomenon. Self-priming has been observed by plotting the velocity vector fields of water. Suction of water in the venturi scrubber occurred due to the difference between static pressure in the venturi scrubber and the hydrostatic pressure of water inside the tank. Dust particle removal efficiency has been calculated for inlet air velocities of 1 m/s and 3 m/s. It has been observed that removal efficiency is higher in case of higher inlet air velocity.