• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.131-141
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• 2021

Simulation is a widely used functional verification method for FBD programs of PLC-based digital I&C system in nuclear power plants. It is difficult, however, to estimate the thoroughness (i.e., effectiveness or quality) of a simulation in the absence of any clear measure for the estimation. This paper proposes two sets of structural coverage adequacy criteria for the FBD simulation, toggle coverage and modified condition/decision coverage, which can estimate the thoroughness of simulation scenarios for FBD programs, as recommended by international standards for functional safety. We developed two supporting tools to generate numerous simulation scenarios and to measure automatically the coverages of the scenarios. The results of our experiment on five FBD programs demonstrated that the measures and tools can help software engineers estimate the thoroughness and improve the simulation scenarios quantitatively.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2017.10a
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• pp.341-342
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• 2017

Control and development systems such as air traffic control systems, road traffic systems, and Korea Hydro &Nuclear Power are the infrastructure facilities of the country, and if the malicious hacking attacks proceed, the damage is beyond imagination. In fact, Korea Hydro & Nuclear Power has been subjected to a hacking attack, causing internal information to leak and causing social problems. In this study, we analyze the environment of the development control system and analyze the status of the convergence security research, which is a recent issue, and propose a strategy system for stabilizing various power generation control systems and propose countermeasures. We propose a method to normalize and integrate data types from various physical security systems (facilities), IT security systems, access control systems, to control the whole system through convergence authentication, and to detect risks through fusion control.

• Proceedings of the KIEE Conference
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• 1998.07c
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• pp.1170-1172
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• 1998

The electrical power of nuclear power plant consists of Safety related power systems and Non-Safety related power systems. The safety related power systems are designed to have sufficient capacity to safely shut down the unit and to mitigate the effects of an accident assuming loss of off-site power. This paper presents the operation scheme of the safety related power system for several plant conditions in Korean Next Generation Reactor and reviews the diversity of power supply to the safety related bus.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Proceedings of the KIEE Conference
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• 2007.11b
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• pp.176-179
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• 2007

This paper proposes a fuzzy linear programming based solution approach fur the long-term generation mix with multi-stages (years) considering air pollution constraints on $CO_2$ emissions, under uncertain circumstances as like as ambiguities of budget and reliability criterion level. This paper approaches to generation mix problem for 2030 year in Korea eventually. The proposed approach may give more flexible solution rather than too robust plan. The effectiveness of the proposed approach is demonstrated by applying it to solve the multi-years best generation mix problem on the Korea power system which contains nuclear, coal, LNG, oil and pumped-storage hydro plants.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.801-813
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• 2023

The events at Three Mile Island in the United States brought about fundamental changes in the ways that simulation would be used in nuclear operations. The need for research simulators was identified to scientifically study human-centered risk and make recommendations for process control system designs. This paper documents the human factors research conducted at the Human Systems and Simulation Laboratory (HSSL) since its inception in 2010 at Idaho National Laboratory. The facility's primary purposes are to provide support to utilities for system upgrades and to validate modernized control room concepts. In the last decade, however, as nuclear industry needs have evolved, so too have the purposes of the HSSL. Thus, beyond control room modernization, human factors researchers have evaluated the security of nuclear infrastructure from cyber adversaries and evaluated human-in-the-loop simulations for joint operations with an integrated hydrogen generation plant. Lastly, our review presents research using human reliability analysis techniques with data collected from HSSL-based studies and concludes with potential future directions for the HSSL, including severe accident management and advanced control room technologies.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.95-103
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• 2014

During the international effort to develop the next generation nuclear reactor technologies, many new power cycle concepts were derived to improve efficiency and reduce the capital cost. Among many innovative power cycles, it was identified that the supercritical $CO_2$ (S-$CO_2$) Brayton cycle technology has a big potential to outperform the existing steam cycle and eventually replace it. The S-$CO_2$ cycle achieves high efficiency with very compact size, which is the ultimate advantage for a power cycle to have. The S-$CO_2$ cycle has a great potential not only for the future nuclear applications but also for general heat sources such as coal, natural gas, and concentrated solar. In this paper, a brief introduction to the S-$CO_2$ power cycle technologies will be first provided, and a short summary of current research and development status of the power cycle technology around the world will be followed. Especially the research works performed by KAIST, KAERI and several related research institutions in Korea will be reviewed in more detail, since they have recently developing a strong infrastructure to test these ideas by constructing a demonstration facility while producing many innovative ideas to improve and realize the concept.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.759-766
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• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

• Journal of Institute of Control, Robotics and Systems
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• v.19 no.4
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• pp.328-333
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• 2013

Since a slight malfunction of control systems in a nuclear power plant may cause huge catastrophes, such control systems usually have multiple redundancy and reliable features, and their reliability and availability should be analyzed and verified thoroughly. This paper performed the reliability analysis of the SPLC (Safety Programmable Logic Controller) that is under developed as the control systems for the next generation nuclear power plant. One of the key features of SPLC is that it has multiple redundancy modes as faults happen, which means the reliability analysis for one fixed redundant model is not enough to analyze the reliability of SPLC. With considering this reconfigurable concept, FTA (Fault Tree Analysis) was used to capture fault-relationship among sub-modules. The analysis results show that MTTF (Mean Time to Fault) of SPLC is 45,080 hours, which is a about 4.5 times longer than the regulation, 10,000 hours.

• Korean Journal of Materials Research
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• v.9 no.2
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• pp.217-225
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• 1999

The effect on the pitting corrosion resistance of laser welding and surface treatment developed as a repair method of stream generator tubing material that was a major component of primary system at nuclear power plant was observed. Some heat-treated Alloy 600 tubing materials used at domestic nuclear power plants were laser-surface observed. Some heat-treated Alloy 600 tubing materials used at domestic nuclear power plants were laser-surface melted and the microstructural characteristics were examined. The pitting corrosion resistance was examined through Ep(pitting potential) and degree of pit generation by means of the electrochemical tests and the immersion tests respectively. The pit formation characteristics were investigated through microstructural changes and the pit initiation site and pit morphology. The test results showed that the pitting corrosion resistances was increased in the order of the followings; sensitized Alloy 600, solution annealed alloy600, and laser surface melted Alloy 600. Pits were initiated preferably at Ti-containing inclusions and their surroundings in all tested specimens and it is believed that higher pitting resistance of laser-surface treated Alloy 600 was caused by fine, homogeneous distribution of non-soluble inclusions, the disappearance of grain boundary, and the formation of dense, stable oxide film. The major element of corrosion products filled in the pit was Cr. On the other hand, Fe was enriched in the deposit formed on the pit.