• The Journal of the Korea institute of electronic communication sciences
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• v.9 no.10
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• pp.1117-1123
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• 2014

Nuclear industries have faced unfavorable circumstances such as an obsolescence of the instrumentation and control system, and therefore nuclear society is striving to resolve this trouble fundamentally. FPGAs are currently highlighted as an alternative means for obsolete control systems. Because of the obsolescence-unaffected characteristics, FPGA should be highly reliable in order to be a replacement for PLC (Programmable Logic Controller). Therefore, it is necessary to establish a software development aspect strategy that enhances the reliability of an FPGA-based controller. The reliability analysis including the MTBF (Mean Time Between Failures) is carried out based on the MIL-HDBK-217F. MTBFs are compared with the FPGA-based controller COMMON-Q PLC. As an analysis result, it shows that the reliability of the FPGA-based controller is better than or equal to that of PLC.

• Journal of Radiation Industry
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• v.17 no.3
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• pp.265-273
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• 2023

To effectively and safely manage the radiation exposure to nuclear power plant (NPP) workers in accidents, major overseas NPP operators such as the United States, Germany, and France have developed and applied realistic 3D model radiation dose assessment software for workers. Continuous research and development have recently been conducted, such as performing NPP accident management using 3D-VR based on As Low As Reasonably Achievable (ALARA) planning tool. In line with this global trend, it is also required to secure technology to manage radiation exposure of workers in Korea efficiently. Therefore, in this paper, it is described the application method and assessment results of radiation exposure scenarios for workers in response to accidents assessment technology, which is one of the fundamental technologies for constructing a realistic platform to be utilized for radiation exposure prediction, diagnosis, management, and training simulations following accidents. First, the post-accident sampling after the Loss of Coolant Accident(LOCA) was selected as the accident and response scenario, and the assessment area related to this work was established. Subsequently, the structures within the assessment area were modeled using MCNP, and the radiation source of the equipment was inputted. Based on this, the radiation dose distribution in the assessment area was assessed. Afterward, considering the three principles of external radiation protection (time, distance, and shielding) detailed work scenarios were developed by varying the number of workers, the presence or absence of a shield, and the location of the shield. The radiation exposure doses received by workers were compared and analyzed for each scenario, and based on the results, the optimal accident response scenario was derived. The results of this study plan to be utilized as a fundamental technology to ensure the safety of workers through simulations targeting various reactor types and accident response scenarios in the future. Furthermore, it is expected to secure the possibility of developing a data-based ALARA decision support system for predicting radiation exposure dose at NPP sites.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2534-2546
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• 2021

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency. Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.

• Journal of Convergence Society for SMB
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• v.2 no.1
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• pp.11-18
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• 2012

As scale of software has been expanded and complicated, it is difficult to detect hazards which induce functional failure of software. Functional failure of safety-critical system (nuclear power plant, air traffic control systems, railway operating system) could result in a disaster (personal injury, environmental pollution). Therefore, it is necessary to conduct a safety analysis for preventing functional failure and increasing safety of the software. However, there are some reasons (time and effort problem, low knowledge of various safety analysis techniques, selecting conventional technique in company, organization) which disturb selecting an apposite one. This paper presents some traditional safety analysis techniques, recently presented techniques and combined models. We expect that it helps stakeholders to choice adequate one for target system.

• Journal of Energy Engineering
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• v.13 no.1
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• pp.51-59
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• 2004

The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited compulsational capability at that time, they usually used very simplified physical models for a real-time simulation of NSSS thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called "negative training", especially for complicated two-phase flows in the reactor coolant system. In resolve the problem, KEPRI developed a realistic NSSS T/H program ARTS which was based on the RETRAN-3D code for the improvement of the Nuclear Power Plant full-scope simulator. The ARTS (based on the RETRAN-3D code) guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there is some possibility of failing to calculate in the case of large break loss of coolant accident (LBLOCA) and low-pressure low-flow transient. In this case, the backup calculation system cover automatically the ARTS. The backup calculation system was expected to provide substantially more accurate predictions in the analysis of the system transients involving LBLOCA. The results were reasonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with FSAR and the AMSI/ANS-3.5-1998 simulator software performance criteria.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1686-1704
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• 2021

This study addresses the numerical simulation of the shield building of an AP1000 nuclear power plant (NPP) subjected to a large commercial aircraft impact. First, a simplified finite element model (F.E. model) of the large commercial Boeing 737 MAX 8 aircraft is established. The F.E. model of the AP1000 shield building is constructed, which is a reasonably simplified reinforced concrete structure. The effectiveness of both F.E. models is verified by the classical Riera method and the impact test of a 1/7.5 scaled GE-J79 engine model. Then, based on the verified F.E. models, the entire impact process of the aircraft on the shield building is simulated by the missile-target interaction method (coupled method) and by the ANSYS/LS-DYNA software, which is at different initial impact velocities and impact heights. Finally, the laws and characteristics of the aircraft impact force, residual velocity, kinetic energy, concrete damage, axial reinforcement stress, and perforated size are analyzed in detail. The results show that all of them increase with the addition to the initial impact velocity. The first four are not very sensitive to the impact height. The engine impact mainly contributes to the peak impact force, and the peak impact force is six times higher than that in the first stage. With increasing initial impact velocity, the maximum aircraft impact force rises linearly. The range of the tension and pressure of the reinforcement axial stress changes with the impact height. The perforated size increases with increasing impact height. The radial perforation area is almost insensitive to the initial impact velocity and impact height. The research of this study can provide help for engineers in designing AP1000 shield buildings.

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.6
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• pp.519-523
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• 2013

A steam generator (SG) tube is an important component of a nuclear power plant (NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC (outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS (computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.

• Journal of KIISE:Computing Practices and Letters
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• v.15 no.4
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• pp.265-269
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• 2009

Formal specification plays important roles in guaranteeing software safety of safety-critical systems such as nuclear power plant's digital control systems. We had developed a technique [1] which synthesizes Function Block Diagram(FBD) programs from NuSCR formal requirements specifications, but it did not be used widely as it had no automatic tool support. FBD is one of the programming languages for Programmable Logic Controllers(PLC) based system. This paper introduces a CASE tool, NuSCRtoFBD, developed to automate the synthesis procedure. The CASE tool NuSCRtoFBD can reduce a number of errors occurred in the process of manual FBD programming.

• Smart Structures and Systems
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• v.13 no.6
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• pp.1065-1094
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• 2014

Civil engineers always face the challenge of uncertainty in planning, building, and maintaining infrastructure. These works rely heavily on a variety of surveying and monitoring techniques. Unmanned aerial vehicles (UAVs) are an effective approach to obtain information from an additional view, and potentially bring significant benefits to civil engineering. This paper gives an overview of the state of UAV developments and their possible applications in civil engineering. The paper begins with an introduction to UAV hardware, software, and control methodologies. It also reviews the latest developments in technologies related to UAVs, such as control theories, navigation methods, and image processing. Finally, the paper concludes with a summary of the potential applications of UAV to seismic risk assessment, transportation, disaster response, construction management, surveying and mapping, and flood monitoring and assessment.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.14 no.1
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• pp.33-38
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• 2013

The general purposed control system for driving a motion of many different typed robot end-effector, which consists of a controller based on ARM Cotex M3-11017 MCU and an application software for generating a motion of end-effector, was developed. Experimental results show that a positioning error is nearly negligible and a repeatability error is 0.04%. Accordingly the developed control system can be applied practically to actuate a robot end-effector for inspection and maintenance of steam generator heat pipe in nuclear power plant.