• Nuclear Engineering and Technology
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• 제11권1호
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• pp.29-45
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• 1979

이 논문에서는 6개의 다른 연료봉 모형을 기준으로 하여 TRIGA Mark-III 연구용 원자로의 핵특성이 자세히 분석되고 있다. 현재 핵연료로 사용하고 있는 FLIP fuel은 연료의 수명을 증진시키기 위해 개발되었던 모형이나, 70%의 고농축이기 때문에 비확산금지조약에 위배되어 수입이 곤란한 실정이다. 따라서 본 논문에서는 FLIP 연료 모형에 대치할 수 있는 다른 연료봉온 찾고 그들의 핵특성을 비교함으로써 앞으로 연료수입계륵을 설립하고자 한다.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2008년도 춘계학술대회논문집
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• pp.641-646
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• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2005년도 추계학술대회 논문집 Vol.18
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• pp.167-168
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• 2005

The operational characteristics of the integrated three-phase flux-lock type superconducting fault current limiter (SFCL) were analyzed. The suggested three-phase SFCL consisted of a three-phase flux-lock reactor and three high-Tc superconducting (HTSC) elements. The former has three windings wound on an iron core, each of which has the same turn's ratio between coil 1 and coil 2. The latter are connected in series with coil 2 of each phase. The integrated three-phase flux-lock type SFCL showed the operational characteristics that the fault phase could affect the sound phase, which resulted in quenching the HTSC element in the sound phase. Through the computer simulation applying numerical analysis for its three-phase equivalent circuit, the fault current limiting characteristics of the integrated three-phase flux-lock type SFCL according to the ground fault types were compared.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2422-2430
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• 2020

In order to improve calculational efficiency of the CAPP code in the analysis of the hexagonal reactor core, we have tried to implement a refined AFEN method with transverse gradient basis functions and interface flux moments in the hexagonal geometry. The numerical scheme for the refined AFEN method adopted here is the response matrix method that uses the interface partial currents as nodal unknowns instead of the interface fluxes used in the original AFEN method. Since the response matrix method is single-node based, it has good properties such as good calculational efficiency and parallel computing affinity. Because a refined AFEN method equivalent nonlinear FDM response matrix method tried first could not provide a numerically stable solution, a direct formulation of the refined AFEN response matrix were developed. To show the numerical performance of this response matrix method against the original AFEN method, the numerical error analyses were performed for several benchmark problems including the VVER-440 LWR benchmark problem and the MHTGR-350 HTGR benchmark problem. The results showed a more than three times speedup in computing time for the LWR and HTGR benchmark problems due to good convergence and excellent calculational efficiency of the refined AFEN response matrix method.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2006년도 추계학술대회 논문집 광주전남지부
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• pp.112-113
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• 2006

The operational characteristics of the integrated three-phase flux-lock type superconducting fault current limiter (SFCL) were analyzed. The suggested three-phase SFCL consisted of a three-phase flux-lock reactor and three high-$T_c$ superconducting (HTSC) elements. The former has three windings wound on an iron core, each of which has the same turn's ratio between coil 1 and coil 2. The latter are connected in series with coil 2 of each phase. The integrated three-phase flux-lock type SFCL showed the operational characteristics that the fault phase could affect the sound phase, which resulted in quenching the HTSC element in the sound phase. Through the computer simulation applying numerical analysis for its three-phase equivalent circuit, the fault current limiting characteristics of the integrated three-phase flux-lock type SFCL according to the ground fault types were compared.

• Nuclear Engineering and Technology
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• 제12권3호
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• pp.153-162
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• 1980

냉각재상실사고의 재관수 단계중 연료봉 피복재의 온도거동 및 열전달 기구를 파악하는 것은 비상노심냉각계통 및 원자로의 안전성해석에 중요하다. 냉각재유동채널의 방위가 rewetting과정에 미치는 영향을 연구하기 위하여 수직 및 수경 유동채널을 이용한 실험을 수행하였으며, 노심이 수평압력관으로 구성되어 있는 CANDU원자로에 관한 실험을 중점적으로 수행하여 그 결과를 수직채널의 결과와 비교 하였다. 또한 rewetting현상을 육안관찰가기 위해 환상형 테스트부 및 외부에서 가열되는 석영관을 사용하였다. 실험결과로써 수평채널에서의 rewetting 속도는 유동의 층상 현상에 크게 영향을 받으나 그 평균값은 수직채널리 경우와 큰차이없음을 알 수 있었다.

• 한국소음진동공학회논문집
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• 제20권3호
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• pp.215-222
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• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
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• 제18권2호
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• pp.71-77
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• 1986

원자로 노심을 축방향으로 일차원 해석을 하고, 가입경수로형원자로의 안전성 해석에 적용할 수 있는 중성자 동특성프로그램 BIK를 개발하였다. BIK프로그램내에서 공간변수에 대해서는 유한차분법이, 시간변수에 대해서는 $\theta$-시간적분법이 채택되었다. 또한 도플러 및 감속재 궤환과 제어봉구동 등을 자세히 묘사하는 모델들이 포함되었다 핵모델의 검증은 ANL검증문제를 통해 이루어졌고, 고리 1호기의 제어봉 인출사고시의 노심출력 변화를 계산하였다. 이상의 계산결과 BIK동특성프로그램이 노심의 중성자 속 변화를 일차원해석의 한계내에서 비교적 정착하게 묘사할 수 있으며, 가압경수로형 원자로의 안전성 해석에 유용하게 사용될 수 있다는 것이 증명되었다.

• 대한기계학회논문집B
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• 제40권12호
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• pp.815-824
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• 2016

가압경수로에 장전되는 핵연료집합체는 연료 봉 다발과 지지격자 및 상하단 고정체로 구성되어 있다. 고온 고압의 냉각수는 원자로 하부로 유입되어 연료 봉 사이로 형성된 부수로를 따라 노심 상부로 흐른다. 경수로핵연료의 주요 열수력 성능인자는 정상운전시 압력강하 및 임계열속이며 사고시에는 급랭 시간이다. 한국원자력연구원에서는 경수로핵연료의 성능을 향상시키고 국산화를 위해 고성능 경수로핵연료, 이중냉각 핵연료 및 사고저항성 핵연료를 개발하였다. 경수로핵연료의 열수력 핵심기술을 개발하기 위해 압력강하 실험, 난류 유동혼합/열전달 실험, 임계열속 및 급랭 시험을 수행하였으며 전산유체역학 방법도 활용하였다. 더불어 사용후핵연료의 임시저장을 위한 건식저장 용기의 열유동에 대한 전산유체해석을 수행하였다. 한편, 경수로핵연료의 열수력 기반기술을 개발하고 실용화를 위해 대학 및 산업체와 협력연구도 진행하였다.