• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1310-1317
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• 2017

GASFLOW-MPI is a widely used scalable computational fluid dynamics numerical tool to simulate the fluid turbulence behavior, combustion dynamics, and other related thermal-hydraulic phenomena in nuclear power plant containment. An efficient scalable linear solver for the large-scale pressure equation is one of the key issues to ensure the computational efficiency of GASFLOW-MPI. Several advanced Krylov subspace methods and scalable preconditioning methods are compared and analyzed to improve the computational performance. With the help of the powerful computational capability, the large eddy simulation turbulent model is used to resolve more detailed turbulent behaviors. A backward-facing step flow is performed to study the free shear layer, the recirculation region, and the boundary layer, which is widespread in many scientific and engineering applications. Numerical results are compared with the experimental data in the literature and the direct numerical simulation results by GASFLOW-MPI. Both time-averaged velocity profile and turbulent intensity are well consistent with the experimental data and direct numerical simulation result. Furthermore, the frequency spectrum is presented and a -5/3 energy decay is observed for a wide range of frequencies, satisfying the turbulent energy spectrum theory. Parallel scaling tests are also implemented on the KIT/IKET cluster and a linear scaling is realized for GASFLOW-MPI.

• Nuclear Engineering and Technology
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• 제19권2호
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• pp.77-84
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• 1987

격납용기 구분방 내에서 정화한 열전달 상관식을 얻기 위해 HDR 실험 자료들을 통계적 인 방법으로 분석하였다. 세가지의 blowdown 시험, 즉 V.42, V.43 and V.44를 통해 얻어진 열전달 자료들이 상관식을 유도하는데 사용되었다. 이미 Uchida에 의해 제안되었던 air-to-steam 질량비는 이 실험에서도 역시 가장 중요한 인자로 입증되었다. 이 연구에서는 Uchida의 열전달 상관식으로 만족되지 않은 실험자료들을, 격납용기 대기와 벽 표면의 온도차의 함수로, 또 대기 압력의 함수로 표시하여 수정하려고 시도하였다. 이 종속성외에도 대기의 난류도와 시간에 따른 인자가 고려되어야 한다. 그러나 HDR 계획에서는 흐름속도의 측정 자료가 부정확하기 때문에 정량적인 관계식의 유도는 힘들다. 다만 열전달 계수와 온도 차이의 관계가 밀접하다는 사실을 밝혔으며 특히 강제순환 조건에서 더욱 이 관계는 명백해짐을 볼 수 있다.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.907-914
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• 2015

The energetic steam explosion caused by contact between the high temperature molten core and water is one of the phenomena that may threaten the integrity of the containment vessel during severe accidents of light water reactors (LWRs). We examined the dependence of steam explosion loads in a typical reactor cavity geometry on selected model parameters and initial/boundary conditions by using a steam explosion simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA). Among the parameters, we put an emphasis on the water pool depth that has significance in terms of accident mitigation strategies including cavity flooding. The results showed a strong correlation between the load and the premixed mass, defined as the mass of the molten material in low void zones (void fraction < 0.75). The jet diameter and velocity that comprise the flow rate were the primary factors to determine the premixed mass and the load. The water pool depth also showed a significant impact. The energy conversion ratio based on the enthalpy in the premixed mass was in a narrow range ~4%. Based on this observation, we proposed a simplified method for evaluation of the steam explosion load. The results showed fair agreement with JASMINE.

• 대한토목학회논문집
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• 제26권1C호
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• pp.1-8
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• 2006

본 연구에서는 중금속으로 오염된 불포화토에 대한 동전기 정화 기법 적용시 발생되는 동전기 현상과 정화 특성을 실험적으로 규명하고자 하였다. 이를 위해 사격장에서 채취한 시료에 대하여 포화도에 따른 일련의 동전기 정화 실험을 수행하였다. 실험 결과 불포화토에서의 납은 주로 확산이중층 내부에서의 직접적인 전기이온이동에 의해 제거되었고, 만약 1V/cm의 전압경사에서 초기 오염농도가 양이온교환능 이하이고 단위토립자당 흡착량이 동일하다면 포화도 증가에 따른 납의 이동효율의 감소로 인하여 제거율은 감소하는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2079-2091
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• 2024

In this paper, a model is developed for calculating pH in the presence of organic impurities due to dissolution of paint and/or continuous injection of organic impurities in the sump. The model is implemented in the AnCheBi code for the analysis of chemical behaviors of the iodine in the containment when the pH changes during a severe accident. Validation of the model is performed with P10T2 and P11T1 experiments carried out by AECL in Canada under the BIP project. Importance analyses of the pH calculation model in the AnCheBi code are then performed with the aforementioned experimental data via Latin hypercube sampling on the reaction coefficients, sensitivity analyses of AnCheBi, and calculation of the correlation coefficients between the reaction coefficients and figure of merits (the pH and the concentrations of the various iodine species). From the importance analyses, we provide the sensitivity of the pH calculation model to the change of pH and the concentrations of the various iodine species and the reaction coefficients related with the dominant phenomena underlying the change of pH and the concentrations of the species.

• 한국농공학회지
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• 제39권3호
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• pp.72-82
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• 1997

In all inelastic deformations time rate effects are always present to some degree. Whether or not their exclusion has a significant influence on the prediction of the material behaviour depends upon several factors. In the study of structural components under static loading conditions at normal temperature it is accepted that time rate effects are generally not important. However metals, especially under high temperatures, exhibit simultaneously the phenomena of creep and viscoplasticity. In this study, elastoplastic and elasto-viscoplastic models include nonlinear geometrical effects were developed and several numerical examples are also included to verify the computer programming work developed here in this work. Comparisons of the calculated results, for the elasto-viscoplastic analysis of an internally pressurised thick cylinder under plane strain condition, have shown that the model yields excellent results. The results obtained from the numerical examples for an elasto-viscoplastic analysis of the Nuclear Reinforced Concrete Containment Structure(NRCCS) subjected to an incrementally applied internal pressure were summarized as follows : 1. The steady state hoop stress distribution along the shell layer of dome and dome wall junction part of NRCCS were linearly behave and the stress in interior surfaces was larger than that in exterior. 2.However in the upper part of the wall of NRCCS the steady state hoop stress in creased linearly from its inner to outer surfaces, being the exact reverse to the previous case of dome/dome-wall junction part. 3.At the lower part of wall of NRCCS, the linear change of steady state hoop stress along its wall layer began to disturb above a certain level of load increase.

• 한국전산구조공학회논문집
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• 제35권2호
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• pp.83-91
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• 2022

원자력발전소 기기 내진설계 및 지진해석은 비연계모델을 대상으로 수행된다. 그러나 이러한 비연계해석은 실제 구조물-기기 간 상호작용 등의 실제 현상을 모사할 수 없기 때문에 연계해석에 비하여 정확하지 못한 결과를 발생시키게 된다는 한계를 가진다. 이러한 배경 아래 이 연구는 실제 원전 격납건물 구조물 및 관련 부계통을 대상으로 질량비와 고유진동수비를 고려하여 지진 연계해석과 비연계해석을 수행하고, 이를 바탕으로 부계통에서의 응답을 비교 분석하였다. 결과적으로 지진 연계해석 결과가 비연계해석 결과보다 대다수 작은 값을 주는 것을 확인하였다. 이러한 결과는 기존 연구인 단순한 연계모델에 대한 해석 결과와 유사하지만, 부계통 응답 차이는 훨씬 더 두드러지게 나타나는 것을 확인하였다. 또한, 이는 지진파의 입력 주파수의 영향보다는 부계통의 설치위치에 영향을 받는 것으로 확인되었다. 마지막으로 비연계 및 연계 지진해석의 차이가 부계통의 질량비가 크고, 고유진동수가 거의 일치하는 영역에서 발생하는 이유는 이 영역에서 주계통과 부계통 동적 상호작용이 크게 나타나기 때문인 것으로 보인다.