• Nuclear Engineering and Technology
• /
• 제47권3호
• /
• pp.267-283
• /
• 2015

The main objective of this study is to investigate experimentally the two-phase flow characteristics in terms of the direct contact condensation of a steam-water stratified flow in a horizontal rectangular channel. Experiments were performed for both air-water and steam-water flows with a cocurrent flow configuration. This work presents the local temperature and velocity distributions in a water layer as well as the interfacial characteristics of both condensing and noncondensing fluid flows. The gas superficial velocity varied from 1.2 m/s to 2.0 m/s for air and from 1.2 m/s to 2.8 m/s for steam under a fixed inlet water superficial velocity of 0.025 m/s. Some advanced measurement methods have been applied to measure the local characteristics of the water layer thickness, temperature, and velocity fields in a horizontal stratified flow. The instantaneous velocity and temperature fields inside the water layer were measured using laser-induced fluorescence and particle image velocimetry, respectively. In addition, the water layer thickness was measured through an ultrasonic method.

• 한국압력기기공학회 논문집
• /
• 제16권1호
• /
• pp.11-21
• /
• 2020

This study aims to develop and advance the evaluation technology for assessing PHWR safety. For this purpose, the complete loss of AC power or station blackout (SBO) was selected as a target accident scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes the main features of the primary heat transport system with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was achieved successfully by running the present model to check out the stable convergence of the key parameters. Subsequently, through the SBO transient analyses two cases with and without the coolant leakage via the PHTS pumps were simulated and the behaviors of the major parameters were compared. The sensitivity analysis on the amount of the coolant leakage by varying its flow area was also performed to investigate the effect on the system responses. It is expected that the results of the present study will contribute to upgrading the evaluation technology of the detailed thermal hydraulic analysis on the SBO transient of the operating PHWRs.

• 에너지공학
• /
• 제24권3호
• /
• pp.96-108
• /
• 2015

피동원자로건물냉각계통(Passive Containment Cooling System; PCCS)은 전원 공급 없이도 원자로건물 내부의 열을 제거하여 그 건전성을 유지시키기 위한 안전설비이다. 본 연구에서는 현재 연구중인 PCCS를 1400 MWe 가압경수형 원전(APR1400)에 설치하는 경우 PCCS 성능을 분석하였다. 분석도구로 계통열수력분석코드 MARS-KS1.3을 사용하였다. PCCS의 성능분석을 위해 APR 1400 표준안전성분석 보고서를 참고하여 원자로건물 내부의 최대압력을 유발하는 사고 시나리오인 저온관 양단 파단사고를 모의하였다. 이 계산에서는 PCCS, 원자로냉각계통 및 원자로건물의 열수력을 동시에 모의하였다. 계산결과를 통해 기존의 원자로건물 살수계통을 대체하여 PCCS가 원자로건물의 건전성을 유지시킬 수 있음을 확인하였다. 또한 PCCS의 성능에 영향을 줄 수 있는 여러 인자를 변경해가며 민감도 분석을 수행하였고 PCCS의 문제점도 확인하였다.

• 전력전자학회:학술대회논문집
• /
• 전력전자학회 2000년도 전력전자학술대회 논문집
• /
• pp.249-252
• /
• 2000

원자력 교육원 #2(KNPEC-2) 시뮬레이터는 1980년도 중반에 웨스팅하우스에 의해 공급되어 계속 사용되어 오다가 현재 성능개선 연구가 진행 중이다. 이번 성능개선을 통해 기존의 컴퓨터 시스템(Gould MPX)와 소프트웨어의 전면 교체가 이루어지고 있으며 최적 계산 코드를 이용한 실시간 열수력 모델 (ARTS; Advanced Real-Time Thermal-Hydraulics Simulation) 개발 , 2-Group 3D 실시간 노심모델(REMARK ; REal Time Multigroup Advanced Reactor Kinetics)를 이용한 노심 주기개선 (Cycle Update) 가상현실 기술 등을 이용한 컴퓨터 교육지원 시스템(CATS: Computer Assister Training System)등 새로운 시도가 이루어지고 있으며 본 논문은 이러한 새로운 시도가 이루어지고 있으며 본 논문은 이러한 새로운 시도들 및 그 결과에 대해 기술하고 있다. 기준발전소(Reference Plant)인 영광 1호기 12주기의 노심모델로 주기개선(Cycle Update)을 위한 REMARK의 입력자료 생성을 위해 핵설계 전산체계인 APA(ALPHA-PHOENIX-ANC) 시스템의 출력으로부터 자동으로 REMARK 입력데이타를 생성하기 위한 GUI툴 개발하였다. 또 이를 이용하여 개발된 노심모델은 최적계산코드(RETRAn 3D) 의 열수력 해법을 이용하여 개발된 NSSS 열수력코드(ARTS) 와 결합(Integration) 되어 안정 및 과도 상태 시험에 사용되었으며 원자로 냉각재 펌프 정지등의 몇 가지 과도 시험 계산결과 기존 해석 결과와 잘 일치하였다 중앙제어실(MCR; Main Control Room)내의 운전원 행동만 훈련하도록 되어있는 기존시뮬레이터의 한계를 극복하기 위해 가상현실 (VR) 저작도구를 이용한 발전소 현장 내부를 표현하는 가상발전소 (Virtual Plant) 발전소 현장에 소재하여 기존 시뮬레이터의 모의한계 밖에 있던 패널을 표현한 가상판넬(Virtual Panel)등과 강의실에서 발전소 모의 훈련을 가능케 하기 위해 가상현실 기술을 이용한 컴퓨터 지원 교육훈력 시스템(CATS ; Computer Assister Training System)을 개발 중이며 일부 개발부분을 소개하였다.

• Nuclear Engineering and Technology
• /
• 제41권7호
• /
• pp.885-892
• /
• 2009

This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid heat conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional "operator split" approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to $1^{st}$ order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant limit required of operator-split methods. In this work, a pilot code was used which employs this tightly coupled, fully implicit method to simulate a reactor core. Results are presented from a simulated control rod movement which show $2^{nd}$ order accuracy in time. Also described in this paper is a simulated rod ejection demonstrating how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.

• Nuclear Engineering and Technology
• /
• 제39권4호
• /
• pp.273-286
• /
• 2007

This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical $CO_2$, an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.

• Nuclear Engineering and Technology
• /
• 제51권4호
• /
• pp.1008-1016
• /
• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• Nuclear Engineering and Technology
• /
• 제56권3호
• /
• pp.980-992
• /
• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

• Nuclear Engineering and Technology
• /
• 제22권1호
• /
• pp.12-19
• /
• 1990

1981년 6월 9일 고리 1호기 원자력발전소에서 발생한 외부 전원 상실사고 자료를 근거로 RELAP5/MOD2코드모델 평가를 하였다. 계산된 주요 열ㆍ수력학 변수를 실측자료와 비교 분석하였으며 증기발생기의 Nodalization 민감도 분석이 수행되었다. 계산된 열ㆍ수력학 변수는 실측치와 비교적 잘 일치하고 있으며, 이러한 유형의 사고 분석에 RELAP5/MOD2가 적합하다는 것을 보였다. 그러나 가압기 압력과 수위변동에서는 상당한 차이를 보였으며 높게 계산되었다. 이러한 사실은 RELAP5의 수직관에서의 층류 열전달 모델에 기인하는 것으로 해당모델의 개선을 요하고 있다는 것을 알았다. 그리고 증기발생기의 Nodalization 연구를 통하여 수위변동을 잘 예측하기 1위해서는 증기발생기 증기 Dome와 Downcomer사이에 압력을 전달시켜주는 유로를 모델링 하여야 한다는 것을 알았다.

• Nuclear Engineering and Technology
• /
• 제41권9호
• /
• pp.1135-1142
• /
• 2009

The activities related to this paper are to investigate the influence of nodalisation on simulating atmospheric stratification in the THAI experiment TH13 (ISP-47) with the German containment code COCOSYS. This article focuses on different nodalisations of the vessel dome, where an atmospheric stratification occurred due to a high helium content. The volume of the dome was divided into several levels that were varied horizontally into different geometries. These geometries differ in the number of zones as well as in the existence of zones that enable the direct rise of an ascending steam plume into the vessel dome. Additionally, the vertical subdivision of the vessel dome was increased to simulate density gradients in a more detailed way. It was pointed out that the proper simulation of atmospheric stratifications and their dissolution depends on both a suitable horizontal as well as vertical nodalisation scheme. Besides, the treatment of fog droplets has an influence if their settlement is not simulated correctly. This report gives an overview of the gained experience and provides nodalisation requirements to simulate atmospheric stratifications and their proper dissolution.