Seoul, South Korea
Seoul, South Korea

Time filter

Source Type

Park J.B.,Korea Radioactive Waste Management Corporation | Kim C.-W.,Korea Hydro and Nuclear Power Co. | Kim S.-H.,FNC Technology | Kim J.Y.,FNC Technology
Annals of Nuclear Energy | Year: 2012

As part of the following-up action pertaining to the construction and operation permit for the first stage of the LILW (Low- and Intermediate-Level Radioactive Waste) repository, the preparation for a large-scale in situ experiment is underway by the Korea Radioactive Waste Management Corporation (KRMC) for a realistic assessment of the characteristics of gas generation after the post-closure phase of a repository. In this paper, we discuss a method of determining the representative composition of simulated dry active waste and a related fabrication plan for this material. After a comparison with experimental gas generation results from Finland, dry active waste content was chosen for a large-scale gas generation experiment. Six different types of materials and details on their simulated dry active waste contents are derived with the total mass and density for both a 200 L and a 320 L drum. © 2011 Elsevier Ltd. All rights reserved.


Jeon S.-S.,FNC Technology | Jeon S.-S.,Seoul National University | Hong S.-J.,FNC Technology | Park J.-Y.,Korea Institute of Nuclear Safety | And 2 more authors.
Nuclear Engineering and Design | Year: 2013

Assessments on the predictive capability of various horizontal in-tube condensation models are performed using the multi-dimensional analysis of reactor safety (MARS) code. In Part I of this two-part paper, the assessments of the stratified flow condensation models were presented while, in Part II, the assessments of the annular flow condensation models are presented. In order to assess the annular flow condensation models, total 19 condensation models were collected from the published literature and incorporated in MARS code, and the heat transfer data was collected from three condensation experiments: JAEA-PCCS, PASCAL, and NOKO experiments. From the assessments, it was found that the present MARS code with combination of the Shah and Chato models under-predicted the condensation heat transfer generally, and the models by Cavallini and Zecchin, Dobson and Chato, Kosky and Staub, Traviss et al., Moser et al., etc., had good predictive capabilities on the condensation heat transfer for the steam-water annular flow. Furthermore, it was found that the stratified flow condensation model by Cavallini et al. (2006), which showed good predictive capability for the Purdue-PCCS experiment (in Part I), was good to use for the prediction of the condensation heat transfer for various steam-water stratified flow conditions. It is expected that the results of this study can be used to significantly improve the condensation models in thermal hydraulic code such as RELAP5 or MARS code. © 2013 Elsevier B.V. All rights reserved.


Jeon S.-S.,FNC Technology | Jeon S.-S.,Seoul National University | Hong S.-J.,FNC Technology | Park J.-Y.,Korea Institute of Nuclear Safety | And 2 more authors.
Nuclear Engineering and Design | Year: 2015

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 essential to analyze and assess the predictive capability of the previous horizontal in-tube condensation models for each flow regime using various experimental data. This study assessed totally 11 condensation models for the stratified flow, one of the main flow regime encountered in the horizontal condenser, with the heat transfer data from the Purdue-PCCS experiment using the multi-dimensional analysis of reactor safety (MARS) code. From the assessments, it was found that the models by Akers and Rosson, Chato, Tandon et al., Sweeney and Chato, and Cavallini et al. (2002) under-predicted the data in the main condensation heat transfer region, on the contrary to this, the models by Rosson and Meyers, Jaster and Kosky, Fujii, Dobson and Chato, and Thome et al. similarly- or over-predicted the data, and especially, Cavallini et al. (2006) model shows good predictive capability for all test conditions. The results of this study can be used importantly to improve the condensation models in thermal hydraulic code, such as RELAP5 or MARS code. © 2012 Elsevier B.V. All rights reserved.


Hong S.J.,FNC Technology | Park G.C.,Seoul National University | Cho S.,Korea Atomic Energy Research Institute | Song C.-H.,Korea Atomic Energy Research Institute
International Journal of Multiphase Flow | Year: 2012

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/m 2s and inversely proportional for over 300kg/m 2s. The frequency was always inversely proportional to pool temperature. For the high steam mass flux region (over 300kg/m 2s), one-dimensional mechanistic model was developed based on the balance of the kinetic energy that the steam jet gives and the pool water receives, adopting the submerged turbulent jet theory. The proposed model excellently predicted the dominant frequencies for the steam mass flux 300-900kg/m 2s and water temperature 35-75°C. For the higher water temperature, the developed model also could predict the dominant frequencies by adjusting the ratio of jet expansion coefficients of vapor dominant region and liquid dominant region. © 2011 Elsevier Ltd.


Kim J.,FNC Technology | Jung H.,Korea Radioactive Waste Management Corporation | Ha J.-C.,Korea Radioactive Waste Management Corporation | Kim E.-H.,Seoul National University
Annals of Nuclear Energy | Year: 2013

In order to simulate gas migration at an underground disposal facility, the characteristics of a medium such as the gas threshold pressure and gas permeability need to be measured in advance. In this study, the gas threshold pressure and gas permeability of silo concrete specimens for a Korean LILW (Low- and Intermediate-Level Waste) disposal facility were measured. The concrete specimens had the same composition as the concrete used in the construction of the silo. The gas threshold pressure was measured by injecting a constant gas flow into cross sections of the specimens. To measure the gas permeability, selected pressures were applied to the specimens and the apparent permeability was calculated using the Hagen-Poiseuille equation. The intrinsic permeability was calculated with the Klinkenberg empirical equation. The gas threshold pressure and gas permeability ranged from 30 to 40 bar and from 10-17 to 10-18 m2, respectively. © 2012 Elsevier Ltd. All rights reserved.


Lee N.R.,FNC Technology | Bang Y.S.,FNC Technology | Lee D.Y.,FNC Technology | Kim H.T.,KHNP Central Research Institute
Annals of Nuclear Energy | Year: 2016

Under severe accidents without containment heat removal, the containment integrity can be challenged due to over-pressurization by the steam and non-condensable gas generation. Containment filtered venting has been considered as an effective measure to avoid or delay the containment failure by over-pressurization. In order to minimize the environmental effects and maximize the effectiveness of filtered venting, it is critical to initiate venting in timely manners with sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to minimize the space and the cost requirement. In this study, the various venting strategies are investigated with respect to possible release flow characteristics. First of all, the accident scenarios representing the core meltdown accident with containment pressurization are selected by reviewing relevant literature. With those scenarios, thermal hydraulic behaviors in the containment and the discharge flow depending on the different venting strategies (i.e. vent line size and vent initiation pressure) are analyzed. MAAP5 model for the OPR1000 Korean nuclear power plant has been used for simulation. © 2016 Elsevier Ltd.


Kim J.,FNC Technology | Jung H.,Korea Radioactive Waste Management Corporation | Ha J.-C.,Korea Radioactive Waste Management Corporation
Nuclear Engineering and Design | Year: 2013

After closure of an underground disposal facility, it would be saturated with ground water and gases would be generated by various mechanisms. As the generated gases are accumulated, internal pressure of the concrete silo would increase and integrity of the silo could be damaged by overpressure. Therefore, an effective gas permeable seal is necessary to prevent the overpressurization of the concrete silo. For designing the gas permeable seal, experimental study and computer modeling on the gas migration through the concrete should be performed. In this study, in order to obtain the specific characteristics of the concrete silo which are necessary for computer modeling, gas entry pressure and gas permeability of the concrete silo were measured by using specimens which have a same composition with the concrete silo. The results of this study could be utilized to the simulation of gas migration and design of the gas permeable seal for a LILW disposal facility in Korea. © 2013 Elsevier B.V. All rights reserved.


Disclosed is a containment filtered venting system (CFVS) for a nuclear power plant, which may include a filtering and venting container which is configured to store the components of the filtered venting system; an inlet pipe which is connected to the filtering and venting container and a reactor building; combined nozzles which are connected to the inlet pipe and are submerged under a filtering solution filled in part of the filtering and venting container; a cyclone separator which is configured to remove larger size substances in droplets and aerosols mixed with the filtering solution from the combined nozzles and guide to a metal filter; a metal filter which is connected to the top of the cyclone separator and is configured to filer impurities mixed in the residual droplets and aerosols; a molecular sieve which is configured to remove organic iodine from exhaust gas filtered by the metal filter; and an outlet pipe which serves to connect the filtering and venting container and a stack.


Provided is an apparatus for testing a loss-of-coolant accident using a model of a nuclear containment building, including a containment vessel of which an upper surface is opened and side and lower surfaces are transparent; an internal structure which is disposed in the containment vessel; a hose pipe which is disposed at an upper side of the containment vessel; and a measuring device which is disposed at the lower surface of the containment vessel so as to monitor movement of fluid and debris in the containment vessel.


PubMed | FNC Technology and Micro Simulation Technology
Type: | Journal: Radiation protection dosimetry | Year: 2016

A conclusion from the lessons learned after the March 2011 Fukushima Daiichi accident was that Korea needs a tool to estimate consequences from a major accident that could occur at a nuclear power plant located in a neighboring country. This paper describes a suite of computer-based codes to be used by Koreas nuclear emergency response staff for training and potentially operational support in Koreas national emergency preparedness and response program. The systems of codes, Northeast Asia Nuclear Accident Simulator (NANAS), consist of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. To quickly assess potential doses to the public in Korea, NANAS includes specific reactor data from the nuclear power plants in China, Japan and Taiwan. The completed simulator is demonstrated using data for a hypothetical release.

Loading FNC Technology collaborators
Loading FNC Technology collaborators