• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.290.2-290
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.267.1-267
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• 1999

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Structural Engineering and Mechanics
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• 제51권6호
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• pp.1005-1016
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• 2014

Roles of reactor internals are to support nuclear fuel, provide insertion and withdrawal channels of nuclear fuel control rods, and carry out core cooling. In case of functional loss of the reactor internals, it may lead to severe accidents caused by damage of nuclear fuel assembly and deterioration of reactor vessel due to attack of fallen out parts. The present study is to examine fluid flows in reactor internals subjected to hydrodynamic loads. In this context, an integrated model was developed and applied to two kinds of numerical analyses; one is to analyze periodic loading effect caused by pump pulsation and the other is to analyze random loading effect employing different turbulent models. Acoustic pressure distributions and flow velocity as well as pressure and temperature fields were calculated and compared to establish appropriate analysis techniques.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.247-247
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• 2002

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.36-43
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• 2023

Lots of researches have been conducted on in-vessel retention (IVR) to prevent or mitigate severe accident in nuclear power plants. Various methodologies were proposed and the external reactor vessel cooling was selected as a part of promising IVR strategy. In this study, the strategy is strengthened by enhancing the natural circulation performance through the adoption of insulation in the reactor cavity. A thermal analysis was carried out based on an assumed accident scenario and its results were used as boundary conditions for subsequent seven flow analysis cases. By comparing the natural circulation performance, effects of annular gaps and insulation shapes on the mass flow rate and flow velocity were quantified. The improvement in cooling performance can be reflected in actual design via detailed assessment.

• Nuclear Engineering and Technology
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• 제27권1호
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• pp.74-83
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• 1995

원자로 압력용기에서의 정화한 속중성자 조사량의 계산은 발전소 압력용기 surveillance program의 핵심적인 부문이다. 최근 기존의 ENDF /B-III~V에 있는 철의 핵단면적 자료가 압력용기와 같은 철이 포함된 구조물에서 속중성자속을 낮게 평가하는 것으로 알려지고 있다. 본 논문에서는 ENDF /B-IV와 VI의 철(Fe) 자료의 비교를 위해 영광3/4호기 모델과 2개의 ENDF/B 파일에 있는 각각의 철자료를 이용하여 47-에너지그룹 핵단면적집 (CXFe-IV와 CXFe-VI )을 만들었다. CXFe-IV와 CXFe-VI를 사용하여 수행한 DOT4.3 계산결과에 의하면 압력용기 취화해석에 중요한 속중성자속(E 〉 1.0 MeV) 계산에서 ENDF /B-VI의 철자료를 사용한 경우가 ENDF /B-IV의 철자료를 사용한 경우보다 압력용기 내부표면에서 7.6%, 외부표면에서 20% 높게 나타났다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.102-107
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• 2004

In general, major components of nuclear power plant have been evaluated based on 2-dimensional design codes conservatively. However, more exact assessment is necessary for continued operation beyond the design life. In this paper, 3-dimensional stress and fatigue analyses reflecting full geometry and monitored operating condition of reactor pressure vessel have been carried out. The analyses results showed that conservatism of current 2-dimensional evaluation based on design transient. Therefore, it is anticipated that the schemes developed from this research such as 3-dimensional finite element modeling, stress analysis and fatigue analysis related techniques can be utilized as fundamental tools for exact lifetime evaluation and license renewal of major nuclear components.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.647-654
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• 2007

Postulation of flaws, one of the most important areas in RPV integrity assessment, significantly affects the results. In the present work, several parameters, such as orientation, underclad vs. surface cracking, crack depth and shape, etc., are postulated and parametric studies are performed to investigate the influence of the flaw parameters on the structural integrity assessment of the reactor pressure vessel during pressurized thermal shock. The influence of individual parameters describing the crack is evaluated based on sensitivity study results.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4551-4559
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• 2022

The aim of this research is to analyze the characteristics of a core melt discharged from the reactor vessel and the spreading behavior the core melt in the reactor cavity of the SMART. First, a severe accident sequence under conservative conditions is simulated by the MELCOR code to obtain the conditions for an analysis of the spreading behavior and coolability of the ex-vessel melt. Second, the spreading behavior and coolability of the ex-vessel melt are analyzed by the MELTSPREAD code. The level, temperature, and pressure of the water in the cavity as well as the temperature, mass, composition, and discharge velocity of the melt were utilized to construct the ex-vessel analysis. The melt spread only to part of the cavity, and that the height of the corium in a static state was less than 25 cm. The characteristics of a small modular reactor on the spreading behavior and coolability of melt were analyzed. In the SMART, the amount of melt discharged into the cavity is relatively small and the area of the cavity is sufficiently large when compared to a high-power pressurized water reactor. It was found that the coolability of an ex-vessel core melt can be sufficiently secured.