• Tunnel and Underground Space
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• v.16 no.1 s.60
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• pp.38-49
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• 2006

LNG storage in lined rock cavern demands various techniques concerned with rock mechanics, thermo-mechanics and hydrogeology in design, construction and maintenance stage. LNG pilot cavern was constructed in Daejeon in order to verify these techniques. In this paper, evaluation of drainage system and ice ring formation was studied by numerical simulation. By Modflow analysis in the viewpoint of aquifer and Seep/W analysis in the viewpoint of flow system, it was verified that the drainage system in the pilot cavern was efficiently operated. Since ice ring formation can be simulated by interactive relation between heat transfer and water flow, coupled analysis of those was performed. In this analysis, the position of ice ring was presumed and it was demonstrated that the formation is affected by velocity and direction of groundwater flow.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1171-1180
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• 2009

The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The $Electricit\acute{e}$ De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code_Saturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.10 no.5
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• pp.535-549
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• 1998

A numerical study for performance analysis of a counterflow type forced draft tower and natural draft cooling tower has been performed based on the method using the finite volume method with non-orthogonal body fitted and non-staggered grid system. For solving the coupling problem between water and air, air enthalpy balance, moisture fraction balance, water enthalpy balance, and water mass balance equations are solved with Navier-Stoke’s equations simultaneously. For the effect of turbulence, the standard k-$\varepsilon$ turbulent model is implied in this analysis. The predicted result of the present analysis is compared with the experimental data and the commercial software result to validate the present study, The predicted results show good agreement with the experimental data and the commercial software result. To investigate the influence of the cooling tower design parameters such as approach, range and wet bulb temperature, parametric studies are also peformed.

• International Journal of Air-Conditioning and Refrigeration
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• v.8 no.1
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• pp.1-13
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• 2000

A numerical study for performance analysis of a crossflow-type forced draft cooling tower has been performed based on the finite volume method with non-orthogonal body fitted, and non-staggered grid system. For solving the coupling problem between water and air, air enthalpy, moisture fraction, water enthalpy, and water mass balance equations are solved with Navier-Stoke's equations simultaneously. For the effect of turbulence, the standard k-$\varepsilon$ turbulent model is implied in this analysis. The predicted result of the present analysis is compared with the experimental data and the commercial software result to validate the present study. The predicted results show good agreement with the experimental data and the commercial software result. To investigate the influence of the cooling tower design parameters such as approach, range and wet bulb temperature, parametric studies are also performed.