• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.3990-4002
• /
• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
• /
• v.29 no.3
• /
• pp.218-228
• /
• 1997

Correlations between the amount of DAW (Dry Active Waste) generated from present Korean PWRs and their operating parameters were analyzed. As the result of multi-variable linear regressions, a model predicting the volume of DAW using the number of shutdowns ( $f_{FS}$ ) and total personnel exposure ( $P_{\varepsilon}$) was derived. Considering one standard error bound, the model could successfully simulate about 8575 of the real data. In order to predict the amount of DAW to be generated from a KNGR another model was derived by taking into account the additional volume reduction by supercompaction system. In addition, the volume of WAW (Wet Active Waste) to be generated from KNGR (Korean Next Generation Reactor) was calculated by considering conceptual design data and replacement effect of radwaste evaporator with selective ion exchangers. Finally, total volume of SRW (Solid Radioactive Waste) to be generated from KNGR was predicted by inserting design goal values of $f_{FS}$ and $P_{\varepsilon}$ into the model. The result showed that the expected amount of SRW to be generated from KNGR would be in the range of 33~44㎥. $y^{-1}$ . It was proved that the value would meet the operational target of KNGR proposed by KEPCO, that is, 50㎥. $y^{-1}$ .

• Nuclear Engineering and Technology
• /
• v.27 no.1
• /
• pp.8-17
• /
• 1995

A great effort has been made to improve the nuclear plant control system by use of digital technologies, and a long term schedule for the control system upgrade has been prepared with an aim to implementation in the next generation nuclear plane. In case of digital control system, it is important to decide the sampling period for analysis and design of the system, because the performance and the stability of a digital control system depend on the value of the sampling period of the digital control system. There is, however, currently no systematic method used universally for determining the sampling period of the digital control system. Generally, a traditional way to select the sampling frequency is to use 20 to 30 times the bandwidth of the analog control system which has the same system configuration and parameters as the digital one. In this paper, a new method to select the sampling period is suggested which takes into account of the performance as well as the stability of the digital control system. By use of the frying's model of steam generator, the optimal sampling period of an assumptive digital control system for steam generator level control is estimated and is actually verified in the digital control simulation system for KORI-2 nuclear power plant steam generator level control. Consequently, we conclude the optimal sampling period of the digital control system for KORI-2 nuclear power plant steam generator level control is 1 second for all power ranges.

• Nuclear Engineering and Technology
• /
• v.52 no.12
• /
• pp.2743-2759
• /
• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• Journal of the Ergonomics Society of Korea
• /
• v.30 no.1
• /
• pp.55-64
• /
• 2011

Objective: The aim of this paper is to introduce the activities and research trends of human reliability analysis including brief summary about contents and methods of the analysis. Background: Various approaches and methods have been suggested and used to assess human reliability in field of risk assessment of nuclear power plants. However, it has noticed that there is high uncertainty in human reliability analysis which results in a major bottleneck for risk-informed activities of nuclear power plants. Method: First and second generation methods of human reliability analysis are reviewed and a few representative methods are discussed from the risk assessment perspective. The strength and weakness of each method is also examined from the viewpoint of reliability analyst as a user. In addition, new research trends in this field are briefly summarized. Results: Human reliability analysis has become an important tool to support not only risk assessment but also system design of a centralized complex system. Conclusion: Human reliability analysis should be improved by active cooperation with researchers in field of human factors. Application: The trends of human reliability analysis explained in this paper will help researchers to find interest topics to which they could contribute.

• Nuclear Engineering and Technology
• /
• v.45 no.3
• /
• pp.401-414
• /
• 2013

HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's $950^{\circ}C$, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to $750^{\circ}C$ for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to $900^{\circ}C$ for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

• Nuclear Engineering and Technology
• /
• v.47 no.6
• /
• pp.678-699
• /
• 2015

Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for nearterm human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of $100MW_{th}$ and an electricity generation mode of $100MW_{th}$, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics was carried out. The result indicates that the innovative design has great potential for high propellant efficiency and thrust-to-weight of engine ratio, compared with the existing NTR designs. However, the build-up of fission products in fuel has a significant impact on the bimodal operation of the moderated reactor such as xenon-induced dead time. This issue can be overcome by building in excess reactivity and control margin for the reactor design.

• Proceedings of the KIEE Conference
• /
• 2005.07a
• /
• pp.149-151
• /
• 2005

A new approach considering CO2 air pollution constraints in the long-term generation mix is proposed under uncertain circumstances. A characteristic feature of the presented approach in this paper is what effects give the air pollution constraints in long term best generation mix. Best generation mix problem is formulated by linear programming with fuel and construction cost minimization with load growth, reliability (reserve margin rate) and air pollutionconstraints. The proposed method accommodates the operation of pumped-storage generator. It was assumed in this study that the construction planning of the hydro power plants is given separately from the other generation plans. The effectiveness of the proposed approach is demonstrated by applying to the best generation mix problem of KEPCO-system, which contains nuclear, coal, LNG, oil and pumped-storage hydro plant multi-years.

• Nuclear Engineering and Technology
• /
• v.52 no.2
• /
• pp.323-327
• /
• 2020

In this study, the neutronic calculation to obtain tritium breeding ratio (TBR) in a deuterium-tritium (D-T) fusion power reactor using Monte Carlo MCNPX is done. In addition, by using COMSOL software, an efficient cooling system is designed. In the proposed design, it is adequate to enrich up to 40% ⁶Li. Total tritium breeding ratio of 1.12 is achieved. The temperature of helium as coolant gas never exceed 687℃. As regards the tolerable temperature of beryllium (650℃), the design of blanket module is done in the way that beryllium temperature never exceed 600℃. The main feature of this design indicates the temperature of helium coolant is higher than other proposed models for blanket module, therefore power of electricity generation will increase.

• Journal of the Ergonomics Society of Korea
• /
• v.18 no.3
• /
• pp.153-169
• /
• 1999

The design of nuclear control room is advancing toward totally computer based human system interfaces (HSI). Computer based interfaces offer the opportunity to provide improved support of operator performance, but if not properly deployed, can introduce new challenges. This paper reviews the Westinghouse AP-600 Human Factors Verification and Validation Plan selected for HSI evaluation model of Korea next generation nuclear control rooms. The AP-600 HSI evaluation model addressed 15 evaluation issues considering major activity class of operator and task complexity factors. This paper also describes the test procedures experimenters should follow to evaluate the addressed issues.