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Park S.H.,Sunchon National University | Park C.,FNC Technology | Lee J.Y.,FNC Technology | Lee B.,FNC Technology
Nuclear Engineering and Technology | Year: 2017

The determination of the shape and rising velocity of gas bubbles in a liquid pool is of great importance in analyzing the radioactive aerosol emissions from nuclear power plant accidents in terms of the fission product release rate and the pool scrubbing efficiency of radioactive aerosols. This article suggests a simple parameterization for the gas bubble rising velocity as a function of the volume-equivalent bubble diameter; this parameterization does not require prior knowledge of bubble shape. This is more convenient than previously suggested parameterizations because it is given as a single explicit formula. It is also shown that a bubble shape diagram, which is very similar to the Grace's diagram, can be easily generated using the parameterization suggested in this article. Furthermore, the boundaries among the three bubble shape regimes in the Eo-. Re plane and the condition for the bypass of the spheroidal regime can be delineated directly from the parameterization formula. Therefore, the parameterization suggested in this article appears to be useful not only in easily determining the bubble rising velocity (e.g., in postulated severe accident analysis codes) but also in understanding the trend of bubble shape change due to bubble growth. © 2017.

Lin C.-S.,Purdue University | Park T.,Purdue University | Park T.,FNC Technology | Yang W.S.,Purdue University
Nuclear Technology | Year: 2017

This paper presents the core design studies of a sodium-cooled fast reactor (SFR) and a sodium-cooled accelerator-driven system (ADS) for a two-stage fast-spectrum fuel cycle to enhance uranium resource utilization and reduce nuclear waste generation. The first-stage SFR starts with low-enriched uranium (LEU) fuel and operates with the recovered uranium and plutonium from the discharged fuels and natural uranium at equilibrium. The recovered minor actinides (MAs) are sent to the second-stage ADS, where they are burned in an inert matrix fuel form. Reference core designs were developed for a 1000-MW(thermal) LEU-fueled breakeven fast reactor (LEUBFR) and an 840-MW(thermal) MA-fueled ADS blanket. The SFR starts with uranium fuel with a 235U enrichment of 13.6% and reaches a fuel-breakeven core after 14 cycles with an 18-month cycle length. At the equilibrium state, one ADS supports 37 fast reactors. Using the performance parameters of SFR and ADS, the proposed two-stage fuel cycle was evaluated. The results of the equilibrium cycle analysis showed that the two-stage fuel cycle option could achieve a high reduction in waste generation because of the continuous recycling of the plutonium and MAs. In addition, the mass flow data showed that this two-stage fuel cycle option increases the efficiency of natural uranium utilization and reduces the nuclear waste generation compared to the conventional two-stage fuel cycle options based on thermal and fast-spectrum systems. © American Nuclear Society.

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: 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.

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.

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