• 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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• 제17권1호
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• pp.8-15
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• 1985

Debris bed내·외로부터 깨스유량을 갖는 debris/water 열적상호작용 해석모델이 중대사고 분석을 위해 제시되었다. 제시된 모델은 증기 소비, debris bed에서 수소 생성, 유입깨스 및 화학반응열에 대한 인자들을 포함하고 있으며, 금속-물반응 및 debris/concrete 작용으로 인한 깨스 생성을 평가하기 위해 MARCH code에 도입시켰다. 그 결과 수소원은 격납용기 과도압력에 큰 영향을 미치나 debris bed로 대류깨스 냉각과 콘크리트로 전도 열손실은 debris bed 냉각성에 조그마한 영향을 주는 것으로 나타났다. 하지만 debris 인자의 재가열과 재용융은 콘크리트와 상호작용에 의해 상당히 지연될 수 있다.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3966-3978
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• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1400-1409
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• 2023

In this study, three condensation models of the CUPID code, i.e., the resolved boundary layer approach (RBLA), heat and mass transfer analogy (HMTA) model, and an empirical correlation, were tested and validated against the COPAIN and CAU tests. An improvement on HMTA model was also made to use well-known heat transfer correlations and to take geometrical effect into consideration. The RBLA was a best option for simulating the COPAIN test, having mean relative error (MRE) about 0.072, followed by the modified HMTA model (MRE about 0.18). On the other hand, benchmark against CAU test (under natural convection and occurred on a slender tube) indicated that the modified HMTA model had better accuracy (MRE about 0.149) than the RBLA (MRE about 0.314). The HMTA model with wall function and the empirical correlation underestimated significantly, having MRE about 0.787 and 0.55 respectively. When using the HMTA model, consideration of geometrical effect such as tube curvature was essential; ignoring such effect leads to significant underestimation. The HMTA and the empirical correlation required significantly less computational resources than the RBLA model. Considering that the HMTA model was reasonable accurate, it may be preferable for large-scale simulations of containment.

• 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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• 제56권1호
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• pp.78-91
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• 2024

Double steel plate concrete composite shear wall (SCSW) has been widely utilized in nuclear power plants and high-rise structures, and its shear connectors have a substantial impact on the seismic performance of SCSW. Therefore, in this study, the mechanical properties of SCSW with angle stiffening ribs as shear connections were parametrically examined for the reactor containment structure of nuclear power plants. The axial compression ratio of the SCSW, the spacing of the angle stiffening rib arrangement and the thickness of the angle stiffening rib steel plate were selected as the study parameters. Four finite element models were constructed by using the finite element program named ABAQUS to verify the experimental results of our team, and 13 finite element models were established to investigate the selected three parameters. Thus, the shear capacity, deformation capacity, ductility and energy dissipation capacity of SCSW were determined. The research results show that: compared with studs, using stiffened ribs as shear connectors can significantly enhance the mechanical properties of SCSW; When the axial compression ratio is 0.3-0.4, the seismic performance of SCSW can be maximized; with the lowering of stiffener gap, the shear bearing capacity is greatly enhanced, and when the gap is lowered to a specific distance, the shear bearing capacity has no major affect; in addition, increasing the thickness of stiffeners can significantly increase the shear capacity, ductility and energy dissipation capacity of SCSW. With the rise in the thickness of angle stiffening ribs, the improvement rate of each mechanical property index slows down. Finally, the shear bearing capacity calculation formula of SCSW with angle stiffening ribs as shear connectors is derived. The average error between the theoretical calculation formula and the finite element calculation results is 8% demonstrating that the theoretical formula is reliable. This study can provide reference for the design of SCSW.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2595-2609
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• 2024

This paper introduces a base isolation system-tuned mass damper inerter (BIS-TMDI) hybrid system to the AP1000 nuclear power plant (NPP), which reduces seismic damage potential of the NPP structure. The effects of fluid-structure interaction (FSI) caused by the passive containment cooling system water storage tank (PCCWST) on NPP's seismic performance are investigated. The FSI of water tank theoretical model is considered based on the Housner's model, and a series of time history analyses are performed to prove the rationality of the proposed model. Three single-objective optimization strategies are employed to minimize the relative displacement variance and absolute acceleration variance of the upper structure, as well as the filtered energy index (FEI). Furthermore, a multi-objective optimization strategy considering all these three indexes is proposed to obtain optimal parameters of vibration control. The influence of vibration control strategies on the relative deformation and acceleration of the upper structure is explored with various water level ratios. The analytical results indicate that the proposed BIS-TMDI strategy has significantly reduced the NPP structure's seismic response. The effectiveness of the vibration control strategy is influenced by the water level ratio, emphasizing the significance of designing an appropriate water level ratio to reduce NPP structure's seismic response.

• 대한토목학회논문집
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• 제10권2호
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• pp.103-112
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• 1990

본 연구는 내진설계시 많은 불확실성을 내포하는 지반의 거동특성을 규명하고, 적용대상지반의 확충 및 경제성 제고를 위하여 토사지반의 동적지반특성 평가 및 지반-구조물 거동 특성을 고찰하였다. 예제해석은 토사지반에 원전 containment 구조물이 설치된 경우를 가상하여 지진하중에 대한 지반-구조물 시스템의 거동을 반무한체해석과 유한요소해석으로 분석하였다. 이는 토사지반에 원전이 건설될 경우에 고려해야 할 안정성 및 경제성 분석의 일환으로 수행되었으며, 토사지반의 큰 비선형거동을 정확하게 해석에 반영하기 위한 해석 software와 지반입력 data의 합리적인 평가방안 등을 예제해석을 통하여 분석하였다. 예제해석결과를 종합해 볼 때 토사지반의 동적거동의 정확한 분석을 위하여 비선형 유한요소해석은 Seed & Idriss 모델이, 선형 유한요소해석은 지진하중에 대한 1차원 지반거동시 변형율에서의 동적지반특성을 이용한 방법이, 반무한체해석은 정적하중시 변형율에서의 동적지반특성을 이용한 방법이 가장 합리적으로 동적지반특성을 평가하는 것으로 추천할 수 있다.

• 한국해양공학회지
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• 제12권1호
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• pp.50-57
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• 1998

The leak before break(LBB) concept is generalized on the design of LNG tanks, pressure vessels and nuclear reactor in that any leakage of containment, in whatever amount, will not result in catastropic failure. For this purpose it is necessary to determine the surface crack shape, the opening displacement and the risk of catastropic brittle fracture when it becomes a through crack. In this study the crack propagation behavior of surface flaws and the crack opening displacement of through cracks under combined membrane and bending stresses were investigated with fatigue tests and fracture toughness test of aluminium alloy A5083-O. And fracture mechanics analysis of the crack opening displacement of through cracks were made in order to develop a new model expressing the behaviors of COD under combined membrane and bending stresses.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1994년도 가을 학술발표회 논문집
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• pp.285-290
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.