• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2006년도 제26회 춘계학술대회논문집
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• pp.130-134
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• 2006

분사면 냉각성능을 개선하기 위한 액체로켓 엔진 연료 매니폴드내의 분리판 설치에 따른 분사균일성 변화를 관찰하였다. 3차원 전산유동해석으로 5개 후보 설계에 대하여 분사균일성을 비교하였으며 최적설계에 대하여 측정결과와 비교함으로써 해석방법을 검증하였다. case I과 II는 매니폴드로 공급되는 유량 전체가 분리판 아래로만 흘러 유속이 크게 증가한다. 하지만 분리판이 끝나는 지점에서의 유속변화와 분사면 중심에서의 유량의 집중으로 분사균일성이 크게 저하된다. 이에 비해 분리판이 입구에서 떨어져 장착된 case III와 IV는 유동이 분리판 위, 아래로 흐를 수 있어 유량집중이 완화되므로 균일한 분사특성을 가진다. 비교한 5가지 설계 중 냉각성능과 분사균일성 측면에서 case IV가 최적으로 판단된다.

• Journal of Mechanical Science and Technology
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• 제16권9호
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• pp.1137-1146
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• 2002

An experiment has been conducted on the flow and heat transfer characteristics of film coolant injected from a row of five holes with compound angle orientations of 35$^{\circ}$ inclination angle and 45$^{\circ}$ orientation angle. The Reynolds number based on the mainstream velocity and injection hole diameter 3.58${\times}$10$^4$. Three-dimensional velocity, film cooling effectiveness and heat transfer coefficient data are presented at three different mass flux ratios of 0.5, 1.0 and 2.0. Flow entrainment has been found between the vortices generated by adjacent injectants. The injectant with compound angle orientation entrains not only the mainstream boundary layer flow but also the adjacent injectant. Because of the flow entrainment, the injectant. With compound angle orientation is characterized by a single vortex while two bound vortices are usually observed in the case of simple angle injection. The strength of the secondary flow depends strongly on the mass flux ratio, which shows significant influence on the film cooling effectiveness and heat transfer coefficient.

• 한국수산과학회지
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• 제18권2호
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• pp.85-93
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• 1985

An experimental investigation on condensing heat transfer for the Refrigerant-11 superheated vapor during condensation on the 40 mm O.D by 75 mm long horizontal tube is carried out under the various conditions of air contents as noncondensable gas, condensing pressure, and coolant temperature. The data span a refrigerant flow range from 23 to 63 kg/h and weight fractions of noncondensable gas range from 0 to $15\%$. The comparisons are made using data obtained by the authors and further data obtained from other sources. The characteristics of the condensing heat transfer of refrigerant superheated vapor with and without noncondensable gas flowing horizontally are revealed experimentally, and on the basis of the data obtained, correlations for predicting heat transfer coefficient during condensation on the tube are proposed.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2001년도 추계 학술대회논문집
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• pp.118-125
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• 2001

The performance prediction of SMART MCP was performed using a computational fluid dynamics code. General capacity-head performance curve of MCP, which is provided to other design branches as design input, was obtained and it showed the typical type of axial pump performance curve. When four MCPs operate in parallel and one of them stops while the others continue to operate, SMART requires reduced power operation. A procedure for predicting the performance of SMART MCP for that case was developed and verified with available experimental data. An analysis based on the developed procedure was performed for two cases; the impeller of sloped MCP is fixed or free to rotate in reverse direction. According to the results, $73\%$ flow rate of normal operation enters the reactor core in the case of the locked impeller. In case of the impeller free rotation, the flow rate entering the reactor core is $62.8\%$.

• 한국자동차공학회논문집
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• 제14권2호
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• pp.39-48
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• 2006

An engine cooling system affects overall performances of an engine which has been recently requested higher power in more confined engine room. The design of efficient cooling system demands a great effort to effectively correlate with each components, such as water jacket, radiator, coolant pump, cooling fan, etc. Thus, the aim of this study is to provide the design tool of the cooling system in the early design stage by enabling for the designer to accurately predict the engine cooling performances. This user-friendly design tool has various ways to assemble each components and control the running condition with related database. The present design tool was simulated and compared with experimental data. As a result, the inlet and outlet temperature of the radiator agree very well with experiments. It was concluded that the present design tool could be effectively used for the design of the engine cooling system.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.109-116
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• 2023

This study presents a CRUD modelling to simulate the thermal resistance behavior of CRUD, deposited on the surface of a cladding tube of a fuel assembly. When heat produced from fuels transfers to a coolant through a cladding tube, the CRUD acting as an additional thermal resistance is expressed as two layers, i.e., a solid oxide layer and an imaginary fluid layer, which are added to the experimental tube's heat structure of the MARS-KS input data. The validation calculation for the experiments performed in UNIST-DISNY facility showed that the center and surface temperatures of the cladding tube increased as the porosity and the steam amount inside pores of the CRUD got higher. In addition, the temperature gradient in the imaginary fluid layer was calculated to be larger than that in the solid oxide part, indicating that the steam amount inside the layer acted more largely as thermal resistance. It was also evaluated through sensitivity calculations that the cladding tube temperature was more sensitive to the CRUD porosity and the steam amount in pores than to the inlet flow rate of the coolant.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1955-1962
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• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.36-48
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• 2022

As a valid numerical method to obtain a high-resolution result of a flow field, computational fluid dynamics (CFD) have been widely used to study coolant flow and heat transfer characteristics in fuel rod bundles. However, the time-consuming, iterative calculation of Navier-Stokes equations makes CFD unsuitable for the scenarios that require efficient simulation such as sensitivity analysis and uncertainty quantification. To solve this problem, a reduced-order model (ROM) based on proper orthogonal decomposition (POD) and machine learning (ML) is proposed to simulate the flow field efficiently. Firstly, a validated CFD model to output the flow field data set of the rod bundle is established. Secondly, based on the POD method, the modes and corresponding coefficients of the flow field were extracted. Then, an deep feed-forward neural network, due to its efficiency in approximating arbitrary functions and its ability to handle high-dimensional and strong nonlinear problems, is selected to build a model that maps the non-linear relationship between the mode coefficients and the boundary conditions. A trained surrogate model for modes coefficients prediction is obtained after a certain number of training iterations. Finally, the flow field is reconstructed by combining the product of the POD basis and coefficients. Based on the test dataset, an evaluation of the ROM is carried out. The evaluation results show that the proposed POD-ROM accurately describe the flow status of the fluid field in rod bundles with high resolution in only a few milliseconds.