• 한국지진공학회논문집
• /
• 제7권2호
• /
• pp.21-27
• /
• 2003

일반적으로 유체-구조물 상호작용을 고려한 유체속 구조물들의 지진 및 진동해석에는 주어진 시스템에 대한 유체부가질량을 추정하여 구조물관 연계하는 단순해석 방법을 주로 사용한다. 실제로 유체속 구조물의 응답특성은 유체부가질량 뿐만 아니라 유체점성으로 인한 감쇠영향을 받으며 이들은 모두 연계항을 갖는 복잡한 행렬 형태로 나타난다. 본 연구에서는 비점성 및 점성 유체에 대한 Navier-Stokes 지배방정식의 선형화를 통한 유한요소 정식화를 유도하였다. 이를 이용하여 유한요소 해석 프로그램을 작성하고 6각형 단면특성을 갖는 액체금속로 노심에 대하여 덕트집합체 사이의 유체간격과 레이놀즈수 변화에 따른 유체부가질량과 유체감쇠에 대한 유한 요소 해석을 수행한 결과, 유체간격이 줄어들수록 유체부가질량은 유체점성의 영향을 크게 받고 유체감쇠는 점성으로 인하여 레이놀즈수의 영향을 크게 받는 것으로 나타났다. 또한 편심을 갖는 동축원통에 대한 유한요소 해석결과, 편심이 증가할수록 유체부가질량은 크게 증가하지만 유체감쇠는 편심이 작은 경우 거의 변화가 없으며 어느 일정 수준이상으로 편심이 커질 경우에는 크게 영향을 받는 것으로 나타났다.

• Nuclear Engineering and Technology
• /
• 제52권8호
• /
• pp.1626-1637
• /
• 2020

This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors. This ambitious benchmark aims to determine the uncertainty in light water reactors systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. The Gesellschaft für Anlagen und Reaktorsicherheit methodology is used to propagate the thermal-hydraulic uncertainty of macroscopic parameters through TRACE5.0p3/PARCSv3.0 coupled code. The main innovative points achieved in this work are i) a new thermal-hydraulic model is developed with a highly-accurate 3D core discretization plus an iterative process is presented to adjust the 3D bypass flow, ii) a control rod insertion occurrence -which data is obtained from a real PWR test- is used as a transient simulation, iii) two approaches are used for the propagation process: maximum response where the uncertainty and sensitivity analysis is performed for the maximum absolute response and index dependent where the uncertainty and sensitivity analysis is performed at each time step, and iv) RESTING MATLAB code is developed to automate the model generation process and, then, propagate the thermal-hydraulic uncertainty. The input uncertainty information is found in related literature or, if not found, defined based on expert judgment. This paper, first, presents the Gesellschaft für Anlagen und Reaktorsicherheit methodology to propagate the uncertainty in thermal-hydraulic macroscopic parameters and, then, shows the results when the methodology is applied to a PWR reactor.

• Nuclear Engineering and Technology
• /
• 제56권8호
• /
• pp.2974-2989
• /
• 2024

The one-dimensional Two-Fluid concept uses an area-average approach to simplify the time and phase-averaged Two-Fluid conservation equations, making it more suitable for addressing difficulties at an industrial scale. Nevertheless, the mathematical framework has inherent weaknesses due to the loss of details throughout the averaging procedures. This limitation makes the conventional model inappropriate for some flow regimes, where short-wavelength perturbations experience uncontrolled amplification, leading to solutions that need to be physically accurate. The critical factor in resolving this problem is the integration of closure relations. These relationships play a crucial function in reintroducing essential physical characteristics, thus correcting the loss that occurs during averaging and guaranteeing the stability of the model. To improve the accuracy of predictions, it is essential to assess the stability and grid dependence of one-dimensional formulations, which are particularly affected by closure relations and numerical schemes. The current research presented in the text focuses on improving the well-posedness of the TFM, specifically within the TRACE code, which is widely utilized for nuclear reactor safety assessments. Incorporating a bubble collision model in the momentum equations is demonstrated to enhance the TFM's resilience, especially in scenarios with high void fractions where conventional TFMs may face challenges. The analysis presents a linear stability analysis performed for the transient one-dimensional Two-Fluid Model of system code TRACE within the framework of vertically dispersed flows. The main emphasis is on evaluating the stability characteristics of the model while also acknowledging its susceptibility to closure relations and numerical techniques.

• Nuclear Engineering and Technology
• /
• 제27권5호
• /
• pp.645-660
• /
• 1995

표준원전을 대상으로하여 저수위 운전시의 잔열제거제통상실사고를 RELAP5/MOD3 및 RELAP5/MOD3.1 전산프로그램을 이용하여 분석하였다. 증기발생기가 이용가능할 때 원자로냉각재계통에 배기 경로가 없는 경우와 배기경로가 있는 경우에 대하여 분석을 수행하였다. 배기경로가 없는 경우에 대해 RELAP5 /MOD3 전산프로그램과 RELAP5 /MOD3.1 전산프로그램으로 비교 분석을 수행하였다. 분석 결과 두 전산프로그램의 계산결과는 정성적인 면 뿐 아니라 정량적 인면도 비교적 잘 일치하였다. 그러나 계산결과로부터 RELAP5 /MOD3의 경우에는 벽 열전달모델의 결함이 발견되어 배기경로가 있는 경우에 대해서는 RELAP5 /MOD3.1 전산프로그램을 이용하여 분석을 수행하였다. 분석결과 원자로정지후 하루가 지났을때 배기경로가 없는 경우에는 두개의 증기발생기로도 잔열이 충분히 제거되지 않아 원자로계통의 압력이 지속적으로 증가하여 사고개시 후4,000초 정도에 원자로계통의 임시밀봉재의 설계압력인 0.24MPa에 도달하였다. 가압기 안전밸브 용량의 세배정도 크기의 배기경로가 있는 경우에는 10,000 초가 지나도 원자로냉자재계통의 압력이 0.24 MPa에 도달하지 않았으며 노심노출이 초래되지 않았다. 분석결과의 상세한 검토를 통해서 저수위 운전시 잔열제거능력 상실사고가 발생하였을 경우 REL-AP5/MOD3.1을 이용한 사고해석 방법론의 타당성을 제안하였으며 또한 적절한 배기용량을 산정하기 위한 자료를 제공하였다.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2007년도 춘계학술대회논문집
• /
• pp.197-201
• /
• 2007

Vibration characteristics of an annular cylindrical fuel rod, which was proposed as a candidate design of fuel's cross section for the ultra-high burnup nuclear fuel, according to the cross-sectional dimensions and the number of supports or the span length were analytically studied. Finite element(FE) modeling for the annular cross sectional fuel was based on the methodology, that have been proven by the test verification, for the conventional PWR nuclear fuel rod. A commercial FEA code, ABAQUS, was used for the FE modeling and analysis. A planar beam element (B21) that uses a linear interpolation was used for the fuel rod and a linear spring element for the spring and dimple of the SG. Natural frequencies and mode shape were calculated according to the preliminary design candidates for the fuel's cross sectional dimension and the number of span. From the analysis results, the design scheme of the annular fuel compatible to the present PWR nuclear reactor core was discussed in terms of the number of supports and fuel's cross section.

• Nuclear Engineering and Technology
• /
• 제9권4호
• /
• pp.223-236
• /
• 1977

가압수형 인자로에 사용되는 이산화우라 핵연료통의 역학적 열적설계 및 성능 분석을 위한 종합적 전산 코드가 개발되었다. PROD 1.0으로 명명된 이 코드에는 연료소자에서 반경 방향으로의 출력 침체, 연료소자의 균열, 고밀화 및 팽창, 핵분열기체의 방출, 피복관의 크립, 냉각수에 의한 열전달 및 부식층의 형성 둥의 제반 현상이 고려되었다. 이 FROD 1.0 코드로써 이차원적 온도 분포, 변형도, 응력 및 피복관 내압 등이 연소시간의 함수로서 적절한 전산 시간이내에 산출된다. 이 코드는 또한 종류가 다른 열중성자로에 쓰이는 산화 연료에도 응용필 수 있다. FROD 1.0의 응용으로서 원자로의 정상가동 상태와 미국 원자력학회 분류의 제 2상태에 해당하는 두 가지의 출력 경로에 더하여, 고리 원자력 발전소 1호기의 초기 노심에 장전된 핵연료봉의 연소특성을 예측하였다. 예측결과는 최종 안전 심사 보고서에 기술된 핵연료봉 설계기준과 비교되었으며 둘 사치의 차이점이 논의되었다.

• Nuclear Engineering and Technology
• /
• 제41권5호
• /
• pp.691-708
• /
• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• 소음진동
• /
• 제5권4호
• /
• pp.515-524
• /
• 1995

Tubes in heat exchanger of fuel rods in reactor core are supported at intemediate point by support p0lates or springs. Current practice is, in case of heat exchanger, to allow clearance between tube and support plate for design and manufacturing consideration. And in case of fuel rod the clearance in support point can be generated due to the support spring force relaxation. Flow-induced vibration of a tube can cause it to impact or rub against support plate or against adjacent tubes and can result in fretting-wear. The tube-to- support dynamic interaction is used to relate experimental wear data from single-span test rigs to real multi-span heat exchanger configurations. The dynamic interaction cna be measured during experimental wear tests. However, the dynamic interaction is difficult to measure in real heat exchangers and, therefore, analytical techniques are required to estimate this interaction. This paper describels the nonlinear impact model of DAGS(Dynamic Analysis of Gapped Structure) code which simulates the tube response to external sinusodial or step excitation and predicts tube motion and tube-to-support dynamic interaction. Three experimental measurements-two single span rods excited by sinusodial force and a two span rod impacted by a steel ball are compared from the simulation nonlinear model of DAGS code. The simulation results from DAGS code are in good agreement with measurements. Therefore, the developed model of DAGS code is good analytical tool for estimating tube-to-support dynamic interaction in real heat exchangers.

• 한국전산유체공학회지
• /
• 제20권4호
• /
• pp.14-20
• /
• 2015

Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.

• Nuclear Engineering and Technology
• /
• 제24권3호
• /
• pp.274-284
• /
• 1992

가압경수로 최적 열수력 분석용 전산코드인 CATHRE의 모델 평가를 위하여 가압경수로의 가상 냉각재 상실사고시 원자로 용기내의 유동현상을 모의한 1/15축소의 CREARE 실험을 모의 계산하였다. 이 실험에서 주요변수들은 비상노심 탱각재 주입량과 아냉정도 그리고 계통압력 및 노심에서 발생되는 증기유량이지만. 본 연구에서는 우선 Downcomer에서 역방향유동의 정성적 분석에 촛점을 맞추었다. 모의 계산 결과와 실험 결과를 비교할 때 정량적인 값 뿐 아니라 변화의 경향에서도 차이가 나타난 것은 주로 적절하지 못한 일부의 수치해석 모델과 상간의 계면마찰 때문으로 판단된다. 따라서 매개변수적 민감도 분석을 통하여 CATHARE 전산코드의‘VOLUME’에 접한 접합점에서 운동량 보존방정식의 상세연구 혹은 다차원 분석을 통해서 이 경우의 물리적 현상을 보다 현실적으로 나타낼 수 있다는 결론을 얻었다.