김응수 교수

The wire-wrapped fuel bundle is an assembly design in a sodium-cooled fast reactor. A wire spacer isused to maintain a constant gap between rods and to enhance the mixing of coolants. The wire makesthe flow complicated by creating a sweeping flow and vortex flow. The vortex affects the flow field andheat transfer inside the subchannels. However, studies on vortices in this geometry are limited. Thepurpose of this research is to investigate the vortex flow created in the wire-wrapped fuel bundle.

The vortex-type fluidic diode (FD) is a key safety component for inherent safety in various advancedreactors such as the sodium fast reactor (SFR) and the molten salt reactor (MSR). In this study, topologyoptimization is conducted to optimize the design of the vortex-type fluidic diode. The optimizationdomain is simplified to 2-dimensional geometry for a tangential port and chamber. As a result, a designwith a circular chamber and a restrictor at the tangential port is obtained. To verify the ne

In a severe accident of light water reactor (LWR), molten core material (corium) can be released into thewet cavity, and a fuel-coolant interaction (FCI) can occur. The molten jet with high speed is broken andfragmented into small debris, which may cause a steam explosion or a molten core concrete interaction(MCCI). Since the premixing stage where the jet breakup occurs has a large impact on the severe accidentprogression, the understanding and evaluation of the jet breakup phenomenon are highly

The Smoothed Particle Hydrodynamics is one of the most widely used mesh-free numerical method forthermo-fluid dynamics. Due to its Lagrangian nature and simplicity, it is recently gaining popularity insimulating complex physics with large deformations. In this study, the 3D single/two-phase numericalsimulations are performed on the Liquid Metal Reactor (LMR) centralized sloshing benchmark experimentusing the SPH parallelized using a GPU. In order to capture multi-phase flows with a large density

This paper proposes strain-based failure model of A533B1 pressure vessel steel to simulate failure, followed by application to OECD lower head failure (OLHF) test simulation for experimental validation. The proposed strain-based failure model uses simple constant and linear functions based on physical failure modes with the critical strain value determined either using the lower bound of true fracture strain or using the average value of total elongation depending on the temperature. Application

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, MA

Quality control in Gas Metal Arc Welding (GMAW) poses challenges due to the complex nature of its underlying processes, making it difficult to identify defects during the welding operations. Conventional post-weld inspections, including non-destructive testing (NDT) methods are inefficient and restrict real-time process optimization. This study introduces a novel deep learning-based multimodal approach for real-time GMAW quality monitoring. The proposed solution integrates electrical data, inclu

Smoothed Particle Hydrodynamics (SPH) is a Lagrangian computational fluid dynamics method that has been widely used in the analysis of physical phenomena characterized by large deformation or multiphase flow analysis, including free surface. Despite the recent implementation of eddy-viscosity models in SPH methodology, sophisticated turbulent analysis using Lagrangian methodology has been limited due to the lack of computational performance and numerical consistency. In this study, we implement

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management (SAM) strategy that has been adopted and used in many nuclear reactors such as AP1000, APR1400, and light water reactor etc. Some reactor accidents have raised concerns about nuclear reactors among residents, leading to a decrease in residents’ acceptability and many studies on SAM are being conducted. Experiments on IVR-ERVC are almost impossible due to its specificity, so fluid characteristic

In this study, we propose a fully parallelized adaptive particle refinement (APR) algorithm for smoothed particle hydrodynamics (SPH) to construct a stable and efficient multi-resolution computing system for nuclear safety analysis. The APR technique, widely employed by SPH research groups to adjust local particle resolutions, currently operates on a serialized algorithm. However, this serialized approach diminishes the computational efficiency of the system, negating the advantages of accelerat

This paper presents the finite element deformation and failure simulation of a typical Korean high-power reactor vessel under a severe accident characterized by large break loss of coolant (LBLOCA) with in-vessel retention of molten corium through external reactor vessel cooling (IVR-ERVC) conditions. Temperature distributions calculated using Modular Accident Analysis Program Version 5 (MAAP5) as thermal boundary conditions were used, and ABAQUS thermal and structural analyses were performed. A

ABSTRACT A steady state Eulerian smoothed particle hydrodynamics (SPH) solver is proposed by integrating the Semi‐Implicit Method for Pressure‐Linked Equations (SIMPLE) algorithm. By removing time dependent terms from the governing equations, the proposed approach directly solves for steady‐state velocity, pressure, and temperature fields for incompressible flows using a matrix‐based formulation. To efficiently handle the resulting large, sparse linear systems, a matrix‐free Bi‐CGSTAB iterative

The in-vessel retention through external reactor vessel cooling (IVR-ERVC) strategy is a key management strategy for early termination of a nuclear severe accident that can threaten the integrity of the reactor vessel. To simulate the physical phenomena of the molten corium, the smoothed particle hydrodynamic (SPH) method is utilized in this study. The SPH method is a Lagrangian computational fluid dynamic (CFD) method that can simulate multi-fluid stratification, turbulence, natural circulation