• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.115.2-115.2
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• 2003

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.655-676
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• 2009

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.19-33
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• 2024

In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as / increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

• Journal of computational fluids engineering
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• v.16 no.2
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• pp.30-48
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• 2011

A three-dimensional thermal-hydraulic code, CUPID, has been developed for the analysis of transient two-phase flows at component scale. The CUPID code adopts a two-fluid three-field model for two-phase flows. A semi-implicit two-step numerical method was developed to obtain numerical solutions on unstructured grids. This paper presents an overview of the CUPID code development and assessment strategy. The governing equations, physical models, numerical methods and their improvements, and the systematic verification and validation processes are discussed. The code couplings with a system code, MARS, and, a three-dimensional reactor kinetics code, MASTER, are also presented.

• 한국전산유체공학회:학술대회논문집
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• 2002.10a
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• pp.64-69
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• 2002

A numerical study has been peformed for evaluation of convection schemes for thermal hydraulic analysis in a liquid metal reactor Four convection schemes, HYBRID, QUICK, SMART and HLPA included in the CFX-4 code are considered. The performances of convection schemes are evaluated by applying them to the five test problems. The accuracy, stability and convergence are tested. It is shown that the HYBRID scheme is too diffusive, and the QUICK scheme exhibits overshoots and undershoots, and the SMART scheme shows convergence oscillations, and the HLPA scheme preserves the boundedness without causing convergence oscillations. The accuracies of SMART, QUICK and HLPA schemes are comparable. Thus, the use of HLPA scheme is highly recommended for thermal hydraulic analysis in a liquid metal reactor.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3902-3909
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• 2021

SACOS series of subchannel analysis codes have been developed by XJTU-NuTheL for many years and are being used for the thermal-hydraulic safety analysis of various reactor cores. To achieve fine whole core pin-level analysis, the input preprocessing and parallel capabilities of the code have been developed in this study. Preprocessing is suitable for modeling rectangular and hexagonal assemblies with less error-prone input; parallelization is established based on the domain decomposition method with the hybrid of MPI and OpenMP. For domain decomposition, a more flexible method has been proposed which can determine the appropriate task division of the core domain according to the number of processors of the server. By performing the calculation time evaluation for the several PWR assembly problems, the code parallelization has been successfully verified with different number of processors. Subsequent analysis results for rectangular- and hexagonal-assembly core imply that the code can be used to model and perform pin-level core safety analysis with acceptable computational efficiency.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.871-894
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• 2013

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This $2^{nd}$ ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.