• 원자력산업
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• 제22권6호통권232호
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• pp.24-32
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• 2002

• 원자력산업
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• 제11권6호통권100호
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• pp.62-63
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• 1991

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.499-504
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• 2001

A validation of a 3D CFD model for predicting local subcooling of moderator in the vicinity of calandria tubes in a CANDU reactor is performed. The small scale moderator experiments performed at Sheridan Park Experimental Laboratory(SPEL) in Ontario, Canada[1] is used for the validation. Also a comparison is made between previous CFD analyses based on 2DMOTH and PHOENICS, and the current model analysis for the same SPEL experiment. For the current model, a set of grid structures for the same geometry as the experimental test section is generated and the momentum, heat and continuity equations are solved by CFX-4.3, a CFD code developed by AEA technology. The matrix of calandria tubes is simplified by the porous media approach. The standard $k-\varepsilon$ turbulence model associated with logarithmic wall treatment and SIMPLEC algorithm on the body fitted grid are used and buoyancy effects are accounted for by the Boussinesq approximation. For the test conditions simulated in this study, the flow pattern identified is a buoyancy-dominated flow, which is generated by the interaction between the dominant buoyancy force by heating and inertial momentum forces by the inlet jets. As a result, the current CFD moderator analysis model predicts the moderator temperature reasonably, and the maximum error against the experimental data is kept at less than $2.0^{\circ}C$ over the whole domain. The simulated velocity field matches with the visualization of SPEL experiments quite well.

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.358-373
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• 1995

중수로용 개량핵 연료집합체인 CANFLEX 핵연료다발의 CANDU-6 원자로 장전시 열수송계통에 대한 유동안정성이 분석되었다. CANFLEX 핵연료다발은 기존의 37개봉 핵연료다발과 원자로출력 및 압력강하 측면에서 거의 일치되며, 이로인해 수력적 거동이 양립하는 반면, CANFLEX핵연료다발은 기존의 37개봉 핵연료다발 보다 임계채널 출력이 증가하며, 반경방향 출력분포의 평탄화로 인해 균일한 엔탈피 분포를 확보할 수 있게 된다. CANFLEX 핵연료다발 및 출구모관들의 상호연결관에 대한 SOPHT 모델을 개발하였으며, 이 모델을 이용하여 CANFLEX 핵연료다발이 장전된 월성 1호기의 유동 안정성 거동이 해석되었다. 해석결과, 열수송계통의 출구모관들의 상호연결관이 없을 경우에는 기존의 37개봉 핵연료다발과 같이 유동이 불안정함을 보였으며, 출구모관들의 상호연결관이 있을 경우에는 정격출력의 $\pm$1% 내에서 안정함을 보였다. 따라서 CANFLEX 핵연료다발의 월성 1호기 장전시 열수송계통의 유동안정성 측면에서는 건전할 것으로 판단되었다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3236-3246
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• 2021

Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.5-6
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• 2005

• 한국가시화정보학회지
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• 제10권3호
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• pp.30-36
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• 2012

We are planning to construct a scaled-down moderator facility to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions. In the present work a preliminary experiment using a 1/40 scaled-down moderator tank has been performed to investigate the anticipated problems of the flow visualization and measurement in the planning scaled-down moderator facility. We shortly describe CFD analysis result for the 1/40 scaled-down test model and the flow measurement techniques used for this test facility under isothermal flow conditions. The Particle Image Velocimetry (PIV) method is used to visualize and measure the velocity field of water in a transparent Plexiglas tank. Planar Laser Induced Fluorescence (PLIF) technique is used to evaluate the feasibility of temperature field measurement in the range of $20-40^{\circ}C$ of water temperature using an one-color method.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

• Journal of Mechanical Science and Technology
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• 제20권7호
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• pp.1030-1042
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• 2006

An experimental study has been performed to investigate the off-take phenomena at the header-feeder systems (horizontal header pipe with multiple feeder branch pipes) in a CANDU (CANadian Deuterium Uranium) reactor with the branch orientation varies ${\pm}36^{\circ}\;or\;{\pm}72^{\circ}$. In order to evaluate the applicability of the conventional correlations used in the safety analysis code, RELAP5-Mod3, the test facility is designed with the 1/2 scale of the. CANDU 6. It was found that the data set for the top, bottom and side branches are in a good agreement with the correlations used. However, for the specific angled branches, ${\pm}36^{\circ}\;and\;{\pm}72^{\circ}$, the onsets of off-take data and quality data showed large deviation with the conventional model inside RELAP5-MOD3. Furthermore, based on the uncertainty analysis, the conventional 2.5 power law needs to be modified. The present experimental data set can be useful for the construction of the general correlation considering the arbitrary branch orientation.