• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• The Journal of the Korea Contents Association
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• v.6 no.11
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• pp.118-125
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• 2006

To solve the reverse function problem of computer games criticized nowadays, the main fact to be concerned in design should be to make feelings of satisfaction during playing and solving the given situation in the game world. In this paper, a design to fulfill the feelings of satisfaction directly and indirectly, controlling playable characters to help another non-playable character in some sort of danger is explained. The requirements to do so is analyzed and the result of demonstration is also presented.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.296-299
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• 2007

KAERI is in the process of carrying out the Nuclear Hydrogen Development and Demonstration (NHDD) Program. The indirect cycle gas cooled reactors that produce heat at temperatures in the order of $950^{\circ}C$ are being considered in the NHDD program. For the indirect gas cooled reactors, the intermediate hear exchanger (IHX) and hot gas duct (HGD) are the main components. For the NHDD program we are in the process of establishing a conceptual design of the IHX and HGD. The pre-conceptual design activities in this study dealt with a preliminary design of the IHX and the HGD including strength and thermal expansion evaluation of the main components.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.103-110
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• 2007

The High Temperature Gas-cooled Reactor (HTGR) possesses inherent safety features and is recognized as a representative advanced nuclear system for the future. Based on the success of the HTR-10, the long-time operation test and safety demonstration tests were carried out. The long-time operation test verifies that the operation procedure and control method are appropriate for the HTR-10 and the safety demonstration test shows that the HTR-10 possesses inherent safety features with a great margin. Meanwhile, two new projects have been recently launched to further develop HTGR technology. One is a prototype modular plant, denoted as HTR-PM, to demonstrate the commercial capability of the HTGR power plant. The HTR-PM is designed as $2{\times}250$ MWt, pebble bed core with a steam turbine generator that serves as an energy conversion system. The other is a gas turbine generator system coupled with the HTR-10, denoted as HTR-10GT, built to demonstrate the feasibility of the HTGR gas turbine technology. The gas turbine generator system is designed in a single shaft configuration supported by active magnetic bearings (AMB). The HTR-10GT project is now in the stage of engineering design and component fabrication. R&D on the helium turbocompressor, a key component, and the key technology of AMB are in progress.

• Proceedings of the KSR Conference
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• 2011.10a
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• pp.3180-3186
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• 2011

A segmented scissors type switch has been developed for the urban transit maglev demonstration line to be commercialized near Incheon International Airport in 2013. Based on the design of the previous segmented 3-way switch, the scissors switch is composed of four segmented 2-way switches up/down and left/right and a turn table in the mid point. The main function of the scissors switch is to change the running direction of the train at end terminals. The developed scissors switch is planned to be installed in front of the 102 station, which has a side platform, of the demonstration line. The total length of the switch is 65m and the distance between the up and down track centerlines is 6m. The 2-way switches and turn table are made of steel box type beams, and have their own driving unit, locking unit, control unit, levitation and propulsion rails, and so on. Installed in the factory, a 100,000-cycle continuous operation test was carried out after manual and automatic test operations. The applicapability of the developed switch was verified through the measurements of the linearity of the track after repetitive operations, the mechanical operation noise, the load of the main driving motor, the safety of the control panel, the natural frequency of the girder, the deformation of the girder, and so on.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• Journal of the Korean Society of Combustion
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• v.15 no.2
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• pp.1-11
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• 2010

Power generation systems based on the oxy-coal combustion with carbon dioxide capture and storage (CCS) capability are being proposed and discussed lately. Although a large number of lab scale studies for oxy-coal power plant have been made, studies of pilot scale or commercial scale power plant are not enough. Only a few demonstration projects for oxy-coal power plant are publicized recently. The proposed systems are evolving and various alternatives are to be comparatively evaluated. This paper presents a proposed approach for performance evaluation of a commercial 100 MWe class power plant, which is currently being considered for 'retrofitting' for the demonstration of the concept. The system is configurated based on design and operating conditions with proper assumptions. System components to be included in the discussion are listed. Evaluation criteria in terms of performance are summarized based on the system heat and mass balance and simple performance parameters, such as the fuel to power efficiency and brief introduction of the second law analysis. Also, gas composition is identified for additional analysis to impurities in the system including the purity of oxygen and unwanted gaseous components of nitrogen, argon and oxygen in air separation unit and $CO_2$ processing unit.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.431-437
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• 2003

The graphic simulator has been used to analyze the problems that can occur during transporting and handling radioactive materials, and to derive necessary devices that can transport and handle spent fuel powder without scattering in a hot cell. The graphic simulator has advantages over the real scale physical mockup with respect to cost and schedule. The process equipment and maintenance devices can be verified in advance with less cost and reduced schedule. The derived results are being reflected in the design of equipment for demonstration and are being verified during demonstration.

• Journal of the Korean Society of Safety
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• v.8 no.4
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• pp.165-174
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• 1993

For the safety assessment of the demonstration-scale incineration plant for treating the combustible radioactive wastes, radiation doses of a worker and a residential individual were estimated. The demonstration plant showed a good performance of trial-burn tests using non-radioactive tracers with resulting In high mass reduction of around 40 times and very low emmission of dusts through a stack, which promised a high decontamination factor in an order of 10$^{7}$ . Based on the result s obtained from the trial-burns in the process, the estimation of radiation dose for workers and general publics near the plant was made using dose pathway calculation theories. The parametric values for calculation were selected from design and operational results of the process and from more conservative conditions In reference data. The estimated annual doses for workers and residential indivisuals were 3.07 $\times$ 10$^{-4}$ and 4.35 X 10$^{-8}$ $\mu$Sv/y, respectively, which were high enough to operate the process when comparing with the allowable dose limit in the regulation. The dose calculation models were quite applicable with showing an excellent safety for the process.

• Journal of the Korean Institute of Gas
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• v.24 no.6
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• pp.83-90
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• 2020

This study conducted a risk assessment using the HyKoRAM program created by international joint research. Risk assessment was conducted based on accident scenarios and worst-case scenarios that could occur in the facility, reflecting design specifications of major facilities and components such as compressors, storage tanks, and hydrogen pipes in the hydrogen charging station, and environmental conditions around the demonstration complex. By identifying potential risks of hydrogen charging stations, we are going to derive the worst leakage, fire, explosion, and accident scenarios that can occur in hydrogen storage tanks, treatment facilities, storage facilities, and analyze the possibility of accidents and the effects of damage on human bodies and surrounding facilities to review safety.