Soon-Joon Hong

• Independent Researcher

The process of pool scrubbing is crucial for managing severe nuclear reactor accidents. However, there are significant uncertainties regarding the retention capacity of fission products. The effectiveness of pool scrubbing has a direct impact on the source term. However, current modeling and prediction capabilities are limited due to conservative a...

The objective of this study is to propose a novel correlation that promises a more accurate representation of bubble size variations in the swarm region at pool scrubbing conditions, requiring only information on the inlet conditions and eliminating the need for complex flow parameter assessments. The recently published experiment studies by Abe et...

Pool scrubbing is an important safety measure for mitigating severe accidents in nuclear reactors, including Small Modular Reactors (SMRs). This process involves complex interactions between hydrodynamic, thermal, and chemical processes. As gas bubbles containing fission products move through a water pool, radioactive substances are effectively tra...

This study investigates Equivalent Spherical Diameter (ESD) estimation at high inlet velocity pool scrubbing conditions using the Interfacial Area Transport Equation (IATE) diameter model including bubble-induced turbulence and interphase modeling. The compatibility of area-averaged Sauter Mean Diameter (SMD), area-averaged Local Equivalent Diamete...

Pool scrubbing is a crucial process in nuclear reactor safety. It involves bubbling contaminated gases through water to remove radioactive particles and reduce environmental release. The Decontamination Factor (DF) measures scrubbing efficiency, defined as the ratio of contaminant mass entering the pool to the post-scrubbing mass. A higher DF indic...

Calculating the bubble diameter accurately is crucial for predicting the De-contamination Factor (DF) in pool scrubbing codes. Currently, empirical correlations that depend on flow regimes and conditions are used to estimate bubble diameters in pool scrubbing codes, leading to a discrepancy between predicted and experimental DFs. The Interfacial Ar...

The effect of chromium coating on zircaloy cladding for LOCA was systematically analyzed using MARS-KS code for Zion plant and FEBA reflood experiment. The equivalent material properties were derived, and they were higher than bare zircaloy cladding, which yielded higher PCT for blowdown phase. Through literature survey, CHF of chromium coated clad...

This paper introduces the CAP version-up to 3.0 and the PT analysis method in order to apply to SMR containment.

The second phase of STELLA program (STELLA-2) has been started to verify and validate the performance of DHRS of PGSFR, and the construction of the facility is expected to be completed in 2019. The STELLA-2 is not only aimed for investigation of decay heat removal characteristics but also demonstrates its performance while interacting with PHTS of...

There have been many efforts to improve the prediction capability of the system codes such as RELAP5, MARS-KS, and SPACE using ATLAS experimental data. However, since the heat loss of the ATLAS has not been reflected correctly in the code input model, there have been differences between the experimental data and the code prediction results. This di...

A horizontal U-shaped heat exchanger (HX) submerged in a pool is under development as a piece of key equipment for a passive safety system in a nuclear power plant (NPP). For the successful design of the HX and the safety analysis of the NPP, reliable prediction of the heat transfer performance of the HX is important. At present, the design and the...

In advanced nuclear power plants, a horizontal U-shaped heat exchanger submerged in a pool is under development as a key equipment of a passive safety system. For the successful design of the heat exchanger and the safety analysis of the nuclear power plant incorporating this passive safety system, the reliable prediction of the nucleate boiling he...

CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equation...

High energy flows by opening of the POSRVs are formed from the pressurizer to the IRWST through the spargers submerged in the IRWST when the RCS is over pressurized. Due to thermal hydraulic loads induced by discharged fluids, it is crucial to understand the phenomena occur in the IRWST and thermal mixing is one of them. This study simulated and an...

A large scale test facility design simulating a pool type sodium fast reactor was carried out on the basis of rigorous scaling approach with a plan of its installation by 2016. In particular, a similarity in heat transfer between solid and fluid were intensively discussed together with the distortion, and viable design parameters were derived. A 1-...

The accurate prediction of the horizontal in-tube condensation heat transfer is a primary concern in the optimum design and safety analysis of horizontal heat exchangers of passive safety systems such as the passive containment cooling system (PCCS), the emergency condenser system (ECS) and the passive auxiliary feed-water system (PAFS). It is esse...

After TMI-2 accident, long-term core cooling management takes more importance rather than short-term management since probabilistic safety assessment performed revealed that long-term management had higher risk than the risk from short-term management. Regarding to this, since 1992, blockage of sump suction strainer was taken a place in Barseback U...

Condensation oscillation of submerged steam jet in water pool was investigated. From the experiments it was found that the dominant frequency of condensation oscillation was proportional to steam mass flux for steam mass flux under 300kg/m2s and inversely proportional for over 300kg/m2s. The frequency was always inversely proportional to pool tempe...

A recirculation sump blockage issue is an important and urgent problem to be resolved for ensuring nuclear power plant safety. Through a series of intensive resolution activities, a new regulatory guide and an evaluation methodology were proposed. However, these were restricted to a conventional pressurized water reactors (PWRs), and the applicabil...

U. S. NRC Regulation Guide 1.52 regulating ESF ACS in nuclear power plants has been revised to revision 3. To apply reduction of operability test time, allowance of alternative challenge agents for in-place leak test of HEPA filters, and upgrade of Methyl Iodide penetration acceptance criterion in activated carbon performance test suggested in Reg....

A nuclear plant ESF ACS simulator was designed, built, and verified to perform experiment related to ESF ACS of nuclear power plants. The dimension of 3D CAD model was based on drawings of the main control room(MCR) of Yonggwang units 5 and 6. The CFD analysis was performed based on the measurement of the actual flow rate of ESF ACS. The air flowin...

In the nuclear power community, deterministic design safety criteria have been used as a major means for assuring safety of nuclear power plants, e.g., light water reactors (LWRs). However, as a result of considerable advances in the quantitative risk analysis technique, such as Probabilistic Risk Assessment (PRA), risk-informed approaches are incr...

Steam generator level measurement is important factor for plant transient analyses using best estimate thermal hydraulic computer codes since the value of steam generator level is used for steam generator level control system and plant protection system. Because steam generator is in the saturation condition which includes steam and liquid together...

This paper discusses a thermal-hydraulic analysis methodology using RETRAN-3D and assembles system analyses for pressurized thermal shock resulting from a steam generator tube rupture accident in Kori Nuclear Unit 1. Through a systematic definition of sequences and thermal-hydraulic analyses using RETRAN-3D, the most important parameters on downcom...

This study conducted mass and energy release experiment for the hot leg large break loss-of-coolant-accident (LBLOCA) during post-blowdown with an integral test facility, Seoul National University Facility (SNUF), and its RELAP5 simulation. This facility simulated the Young Kwang Nuclear Power Plant Units 3 and 4 (YGN3&4) with volume ratio of 1:114...

A thermal-hydraulic analysis methodology is established for a pressurized thermal shock (PTS) analysis of the Kori Nuclear Unit-1 (KNU-1) power plant using RETRAN-3D. The effects of the important parameters on PTS are evaluated, such as the initial power level, break size, isolation of the depressurized steam generator (SG), and charging flow runba...

Hot leg break LBLOCA(Large Break LOCA) had a potential to be a containment maximum pressure accident in YGN3&4, which was induced from excessive conservatism in the CE analysis methodology of mass and energy release. This study conducted mass and energy release experiment for the hot leg break LBLOCA during post blowdown with an integral test facil...