• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.27-37
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• 1986

The neutron fluxes and dose rates due to radiation streaming from reactor cavities were evaluated at the KNU-1 reactor pressure vessel (RPY) head flange elevation. To find a suitable cross section data set for the evaluation, a benchmark test was performed for three data sets; DLC-23/CASK, DLC-31/FEWG, and DLC-47/BUGLE. The leakage fluxes from the KNU-1 RPV outer surface were calculated with two different methods: 1-D calculation with ANISN, and 2-D calculation with DOT3.5. The Monte Carlo procedures as embodied in the MORSE-CG code combined with the albedo option were applied to predict the radiation distributions in the cavity region. Finally, the activation analysis of the stud bolts was performed to identify the major activation products.

• Journal of the Korean Vacuum Society
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• v.17 no.5
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• pp.381-386
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• 2008

Adatom kinetics on the surfaces of Co overlayers, prepared on the W(110) surface, was studied with scanning tunneling microscopy. By counting the number-density of the adatom-islands, we estimated the ratio of adatom diffusion coefficients. The ratio $D_{W(110)}:D_{1ML\;Co}:D_{2ML\;Co}$ was measured to be 1 : 125 : 33000 at room temperature, where $D_{W(110)},\;D_{1ML\;Co}$, and $D_{2ML\;Co}$ are the diffusion coefficients on bare W(110) surface, on one-monolayer Co overlayer, and on two-monolayers Co overlayers, respectively. An increased diffusion coefficient on two-ML Co overlayers, relative to that on one-ML Co overlayers, was explained with the heteroepitaxial strain effect.

• Proceedings of the Korean Institute of IIIuminating and Electrical Installation Engineers Conference
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• 1990.10a
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• pp.13-16
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• 1990

균일한 positive column의 상세한 이론이 충돌 단면적 곡선과 공진 방사 그리고 다단계 여기(multistage)와 이온화 과정을 고려하여 0.5에서 5mmHg 압력에서 수은과 희유가스(rare-gas)가 혼합되어 있는 D.C.형광 램프 방전에 대해 전개된다. 온도와 이온, 비여기원자, 여기원자의 밀도가 수치해석에 의하여 계산된다.

• Journal of the Earthquake Engineering Society of Korea
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• v.4 no.1
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• pp.51-62
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• 2000

핵연료집합체 검증 프로그램의 일환으로 본 연구에서는 지진과 배과파단이 핵연료집합체의 건선성에 미치는 영향을 검토하였다 원자로 노심의 상세 동적해석을 이용하여 지진 및 배과파단시 핵연료 집합체에 발생하는 전단력 굽힘 모우멘트 및 변위를 계산하였고 또한 집합체를 지지하고 있는 지지격자체의 충격력을 검토하였다 이들 하중에 대한 핵연료집합체의 응력해석을 수행하여 사고조건하에서의 구조적 건전성에 대하여 언급하였고 추후 설계시 고려할 사항을 제시하였다.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.16 no.2
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• pp.153-163
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• 2003

Performed here is a comparative assessment study for the generation of the pressure-temperature limit curve of the reactor vessel. A round robin problem is proposed using the data available in Korea and all organizations interested in the generation of the pressure-temperature limit curve are invited. The problems consisting of 12 cases for cool-down are solved and their results are compared to generate a reference solution for the reference problem, which will be useful in the evaluation of the generation of the pressure-temperature limit curve in the future.

• Journal of the Korean Chemical Society
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• v.53 no.3
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• pp.355-367
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• 2009

In this study, the 9th grade students were interviewed to explore their comprehension and interpretation processes of inscriptions representing the concept of atom and molecule. We used a semiotic model for the interview and the analysis of result. The research result revealed that the students performed structuring processes of interpreting inscriptions successfully, but they had a difficulty with translating processes for attaining the target concept or for connecting an inscription with another inscription. Translating processes of connecting inscription with text showed a different result according to achievement level of each student. On the other hand, all the interviewees showed similar sequences in the processes of interpreting inscriptions. Educational implications of these findings were discussed.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.9
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• pp.855-862
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• 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.188-194
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• 1996

현재 국내에서 관심이 고조되고 있는 원전 연장운전 방안의 일환으로, 주요 기기의 기술적 타당성 검토 대상기기 중 하나인 원자로 냉각재 배관 노즐의 피로수명 평가를 수행하였다. 본 노즐의 피로수명 평가는 원자로 냉각재 계통 노즐의 최종 설계문서에 제시된 응력해석 결과를 참조하여 ASME Code, Sec III의 절차에 따라 수행하였으며, 평가결과 이들 노즐의 설계수명 30년을 향후 40년 또는 그 이상 연장 운전할 경우에도 무리가 없는 것으로 판단된다.

• 한국전산유체공학회:학술대회논문집
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• 2009.04a
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• pp.290-297
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• 2009

A two-phase (gas and liquid) flow analysis solver, named CUPID, has been developed for a realistic simulation of transient two-phase flows in light water nuclear reactor components. In the CUPID solver, a two-fluid three-field model is adopted and the governing equations are solved on unstructured grids for flow analyses in complicated geometries. For the numerical solution scheme, the semi-implicit method of the RELAP5 code, which has been proved to be very stable and accurate for most practical applications of nuclear thermal hydraulics, was used with some modifications for an application to unstructured non-staggered grids. This paper is concerned with the effects of interpolation schemes on the simulation of two-phase flows. In order to stabilize a numerical solution and assure a high numerical accuracy, the second-order upwind scheme is implemented into the CUPID code in the present paper. Some numerical tests have been performed with the implemented scheme and the comparison results between the second-order and first-order upwind schemes are introduced in the present paper. The comparison results among the two interpolation schemes and either the exact solutions or the mesh convergence studies showed the reduced numerical diffusion with the second order scheme.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.284-289
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• 1996

국산 원자로 압력용기강(ASME SA508 cl.3)을 대상으로 표준 샤피충격시험편(2 mm V- notch)과 피로균열(precracked Charpy) 시험편을 제작하여 계장화(instrumented)충격시험을 실시하고, 충격시험시 하중점(load point)의 변위(displacement) 혹은 시간의 변화를 하중의 변화와 함께 측정하였다. 측정결과를 파괴현상 및 파괴역학과 연계시켜 해석하므로서, 가능한한 소량의 시험편(혹은 시험공정)을 사용하여 필요로 하는 인성평가 관련 정보를 획득할 수 있도록 시도하였다. 그 결과, 파괴과정을 나타내는 하중의 변화를 이용하여 Shear fraction 을 예측할 수 있었고, 하중의 변화와 관계된 변위로부터 Lateral expansion을 추정할 수 있었다. 피로균열 시험편 시험결과로 부터는 충격시의 항복하중, 항복변위, 최고하중 등을 획득하여 균열크기의 함수로 표시되는 시험편 Compliance 를 계산하였고, Equivalent energy 법과 J-integral 법을 적용하여 원자로 압력용기강의 탄소성 동적파괴인 성을 평가할 수 있었다.