• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1540-1555
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.103-106
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• 2002

This paper presents an effective numerical method for analyzing three-dimensional unsteady conjugate heat transfer problems of a curved pipe subjected to infernally thermal stratification. In the present numerical analyses, the thermally stratified flows in the pipe are simulated using the standard $k-{\varepsilon}$turbulent model and the unsteady conjugate heat transfer is treated numerically with a simple and convenient numerical technique. The unsteady conjugate heat transfer analysis method is implemented in a finite volume thermal-hydraulic computer code based on a non-staggered grid arrangement, SIMPLEC algorithm and higher-order bounded convection scheme. Numerical calculations have been performed far the two cases of thermally stratified pipe flows where the surging directions are opposite each other i.e. In-surge and out-surge. The results show that the present numerical analysis method is effective to solve the unsteady flow and conjugate heat transfer in a curved pipe subjected to infernally thermal stratification.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.518-531
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• 1995

A study has been Peformed to investigate the thermal margin increase by replacing the single-channel analysis model with a multichannel analysis model. h new critical heat flux(CHF) correlation, which is applicable to a 17$\times$17 Korean Fuel Assembly(KOFA)-loaded core, was developed on the basis of the local conditions predicted by the subchannel analysis code, TORC. The hot sub-channel analysis was carried out by using one-stage analysis methodology with a prescribed nodal layout of the core. The result of the analysis shooed that more than 5％ of the thermal margin can be recovered by introducing the TORC/KRB-1 system(multichannel analysis model) instead of the PUMA/ERB-2 system(single-channel anal)sis model). The thermal margin increase was attributed not only to the effect of the local thermal hydraulic conditions in the hot subchannel predicted by the code, but also to the effect of the characteristics of the CHF correlation.

• Journal of Energy Engineering
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• v.24 no.1
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• pp.149-163
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• 2015

In this study, we conducted study on the confirmation of thermal-hydraulic safety for Mast assembly with Computational Fluid Dynamics(CFD) analysis. Before performing the natural convection analysis for the Mast assembly by using CFD code, we validated the CFD code against two benchmark natural convection data for the evaluation of turbulence models and confirmation of its applicability to the natural convection flow. From the first benchmark test which was performed by Betts et al. in the simple rectangular channel, we selected standard k-omega turbulence model for natural convection. And then, calculation performance of CFD code was also investigated in the sub-channel of rod bundle by comparing with PNL(Pacific Northwest Laboratory) experimental data and prediction results by MATRA and Fluent 12.0 which were performed by Kwon et al.. Finally, we performed main natural convection analysis for fuel assembly inside the Mast assembly by using validated turbulence model. From the calculation, we observed stable natural circulation flow between the mast assembly and pool side and evaluated the thermal-hydraulic safety by calculating the departure from nucleate boiling ratio.

• Journal of the Korean Society of Systems Engineering
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• v.19 no.2
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• pp.46-58
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• 2023

Deterministic safety analysis is a crucial part of safety assessment, particularly when it comes to demonstrating the safety of nuclear power plant designs. The traditional approach to deterministic safety analysis models is to model the nuclear core using point kinetics. However, this simplified approach does not fully reflect the real core behavior with proper moderator and fuel reactivity feedbacks during the transient. The use of Multi-Physics approach allows more precise simulation reflecting the inherent three-dimensionality (3D) of the problem by representing the detailed 3D core, with instantaneous updates of feedback mechanisms due to changes of important reactivity parameters like fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC). This paper addresses a CEA ejection accident at hot full power (HFP), in which the underlying strong and un-symmetric feedback between thermal-hydraulics and reactor kinetics exist. For this purpose, a multi-physics analysis tool has been selected with the nodal kinetics code, 3DKIN, implicitly coupled to the thermal-hydraulic code, RELAP5, for real-time communication and data exchange. This coupled approach enables high fidelity three-dimensional simulation and is therefore especially relevant to reactivity initiated accident (RIA) scenarios and power distribution anomalies with strong feedback mechanisms and/or un-symmetrical characteristics as in the CEA ejection accident. The Systems Engineering approach is employed to provide guidance in developing the work in a systematic and efficient fashion.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.10a
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• pp.245-253
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• 2004

There has been recent demand for extending the life of age-degraded structures and equipment by such techniques as diagnosis, maintenance, safety assessment, and estimating residual life on iron-making plants and hydraulic, thermal, and nuclear power plants. These techniques take into account comprehensive scenarios that may cause malfunction and structural damage and allow an assessment of risk based on the likely scenarios. In particular the safety assessment and residual life estimation of age-degraded ships and equipment facilities require consideration of various factors such as mechanical and thermal stresses, corrosion, hardness, load variation due to changes of operating condition, crack generation and strength reduction of material by fatigue. In this study, a detail thermal stress analysis, one of useful techniques of safety assessment and maintenance, is performed on a blast furnace by using general FEM code (MSC/NASTRAN).

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.311-322
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• 2012

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.627-628
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• 2005

• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• 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.