• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3526-3539
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• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to simulate nuclear reactor systems, which solve simplified governing equations by replacing source terms with constitutive relations for simulating entire reactor systems with low computational resources. For half a century, many efforts have been made for wider versatility and higher accuracy of system codes, but various factors can affect the code analysis results, and it was difficult to isolate these factors and interpret them individually. In this study, two system codes, RELAP5 and TRACE, which have many users and are highly reliable, are selected to analyze only the effects of constitutive relations. The influence of constitutive relations is analyzed using in-house platforms that replicate constitute relations of RELAP5 and TRACE equally to exclude factors that may affect analysis results, such as governing equation solvers and user effects. Among the various constitutive relations, the analysis is performed on the wall variables expected to have the most influence on the analysis results. Part 1 paper presents the methodology and wall friction model comparison, while Part 2 paper shows wall heat transfer comparison of the two selected codes.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.131-143
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• 2023

Constitutive equations in a nuclear reactor safety analysis code are mostly empirical correlations developed from experiments, which always accompany uncertainties. The accuracy of the code can be improved by modifying the constitutive equations fitting wider range of data with less uncertainty. Thus, the sensitivity of the code with respect to the constitutive equations is evaluated quantitatively in the paper to understand the room for improvement of the code. A new methodology is proposed which first starts by dividing the thermal hydraulic conditions into multiple sub-regimes using self-organizing map (SOM) clustering method. The sensitivity analysis is then conducted by multiplying an arbitrary set of coefficients to the constitutive equations for each sub-divided thermal-hydraulic regime with SOM to observe how the code accuracy varies. The randomly chosen multiplier coefficient represents the uncertainty of the constitutive equations. Furthermore, the set with the smallest error with the selected experimental data can be obtained and can provide insight which direction should the constitutive equations be modified to improve the code accuracy. The newly proposed method is applied to a steady-state experiment and a transient experiment to illustrate how the method can provide insight to the code developer.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.11-32
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• 2006

System thermal-hydraulic codes have been used in the past decades in the areas of design, operation, licensing and safety of Nuclear Power Plants (NPPs). The development and validation of these codes have reached a high degree of maturity, through the consideration of huge experiments and advanced numerical models. Nowadays, the analyses are based upon realistic approaches rather than the conservative evaluation models. However the applications of these computational tools require preliminary qualification issues. Although huge amounts of financial and human resources have been invested for the development and improvement of codes, the calculation results are still affected by errors. In the sophisticated nuclear technology, design and safety of NPP, these errors must be quantified. An overview of the state of the art of the current thermal-hydraulic system code is developed and the need of uncertainty analysis in code calculations is emphasized. Several sources of uncertainty have been classified and commented, and typical applications of such methods are shown.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.39-42
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• 2002

The nuclear transmutation technology to Incinerate the long lived radioactive nuclides and produce energy during the incineration process is believed to be one or the best solutions. HYPER(${\underline{HY}}brid {\underline{P}}ower {\underline{E}}xtraction {\underline{R}}$eactor)is the accelerator driven transmutation system which is being developed by KAERI(Korea Atomic Energy Research Institute). Lead-bismuth(Pb-Bi) is adopted as a coolant and spallation target material. In this paper, we performed the thermal-hydraulic analysis of HYPER target using the commercial code FLUENT, and also calculated thermal and mechanical stress of the beam window using the commercial code ANSYS. It is found that there is an optimum value for the window diameter and the maximum allowable beam current can be increased to 17.3 mA for the inner diameter of windows, 40 cm. Finally, the other shapes such as uniform or scanned beam were considered. The results of FLUENT calculations show that the uniform type is preferable to the other shapes of the beam in terms of the window and target cooling and the maximum window temperature is lower than that of the parabolic beam by $58 ^{\circ}C$ for the beam current, 13 mA.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.225-236
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• 1994

The sudden discharge of the loop seal water, which is present upstream of the SEBIM POSRV, creates large momentum and inertia forces on the downstream of the discharge piping system. This study provides the procedures and results of analysis of the thermal-hydraulic transient in the SEBIM POSRV discharge piping during the valve opening. The analysis is peformed by RELAP5/MOD3. The appropriate modeling of the discharge piping system, SEBIM POSRV opening characteristics, and loop seal water discharge for the RELAP5/MOD3 analysis is suggested. Also performed is the sensitivity study for the selection of proper options for the junction and volume control. flags. The analysis results demonstrate the adequacy of the RELAP5/HOD3 for the thermal-hydraulic transient analysis of the loop seal water discharge of the SEBIM POSRV discharge piping system. From the sensitivity analysis results, it is shown that the smooth area change option with reasonable geometric pressure drop distribution, non-equilibrium option, and proper time step should be selected for loop seal water discharge analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.427-428
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• 2004

• Transactions of The Korea Fluid Power Systems Society
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• v.4 no.2
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• pp.28-33
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• 2007

In the power plant facility which use soft coal as a power source the fan pitch blade control hydraulic actuator is used to control the inlet and outlet gas to regulate the internal pressure of the furnace and control the frequence. Sometimes malfunctions of this equipment lead to the decline of boiler thermal efficiency and unexpected power plant trip. In order to localize the fan pitch blade control hydraulic actuator specially for the 500MW large scale boiler, Analysis and modelling of the system is carried out mathematically. The responses of the system are examined by using matlab simulation fur the variation of the major parameters in view of reverse engineering. Consequently the validity of the established parameters are examined.

• Journal of the Korean Geotechnical Society
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• v.29 no.8
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• pp.65-73
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• 2013

Ground-coupled heat pump system has attracted attention as a promising renewable energy technology due to its improving energy efficiency and eco-friendly mechanism for space cooling and heating. Pipes buried in the ground play a role of direct thermal interaction between circulating fluid inside the pipe and surrounding soils in the geothermal exchange system. However, both complexities of turbulent flow coupling thermal-hydraulic phenomena and very long aspect ratio of the pipe make it difficult to model the heat exchange system directly. Energy balance for fluid flow inside the pipe was derived to model thermal-hydraulic phenomena, and one-dimensional pipe element was proposed through Galerkin formation and time integration of the equation. Developed element is combined to pre-developed FEM code for THM phenomena in porous media. Numerical results of Thermal Response Test showed that line-source model overestimates equivalent thermal conductivity of surrounding soils due to thermal interaction between adjacent pipes and finite length of the pipe. Thus, inverse analysis for the TRT simulation was conducted to present optimal transformation matrix with utmost convergence.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.