• Nuclear Engineering and Technology
• /
• v.50 no.3
• /
• pp.478-487
• /
• 2018

Pyroprocessing technology has been regarded as a promising solution for recycling spent fuel in nuclear power plants. The Korea Atomic Energy Research Institute has been studying the current status of equipment and facilities for pyroprocessing and found that existing facilities are manually operated; therefore, their applications have been limited to laboratory scale because of low productivity and safety concerns. To extend the pyroprocessing technology to a commercial scale, the facility, including all the processing equipment and the material-handling devices, should be enhanced in view of automation. In an automated pyroprocessing facility, a supervised control system is needed to handle and manage material flow and associated operations. This article provides a preliminary design of the supervising system for pyroprocessing. In particular, a manufacturing execution system intended for an automated pyroprocessing facility, named Pyroprocessing Execution System, is proposed, by which the overall production process is automated via systematic collaboration with a planning system and a control system. Moreover, a simulation-based prototype system is presented to illustrate the operability of the proposed Pyroprocessing Execution System, and a simulation study to demonstrate the interoperability of the material-handling equipment with processing equipment is also provided.

• Nuclear Engineering and Technology
• /
• v.54 no.4
• /
• pp.1363-1373
• /
• 2022

The recent seismic events that occurred in South Korea have increased the interest in the re-evaluation of the seismic capacity of nuclear power plant (NPP) equipment, which is often conservatively estimated. To date, various approaches-including the Bayesian method proposed by the United States (US) Electric Power Research Institute -have been developed to quantify the seismic capacity of NPP equipment. Among these, the Bayesian approach has advantages in accounting for both prior knowledge and new information to update the probabilistic distribution of seismic capacity. However, data availability and region-specific issues exist in applying this Bayesian approach to Korean NPP equipment. Therefore, this paper proposes to construct an earthquake experience database by combining available earthquake records at Korean NPP sites and the general location of equipment within NPPs. Also, for the better representation of the seismic demand of Korean earthquake datasets, which have distinct seismic characteristics from those of the US at a high-frequency range, a broadband frequency range optimization is suggested. The proposed data construction and seismic demand optimization method for seismic capacity re-evaluation are demonstrated and tested on a 480 V motor control center of a South Korea NPP.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.1150-1151
• /
• 2005

• Proceedings of the Earthquake Engineering Society of Korea Conference
• /
• 2003.09a
• /
• pp.411-418
• /
• 2003

This paper presents the results of experimental studies on the equipment isolation effect in the nuclear containment. for this Purpose, shaking table tests were performed. The natural rubber bearing (NRB) and high damping rubber bearing (HDRB) were selected for the isolation. Peak ground acceleration, damping characteristics of isolation system and frequency contents of selected earthquake motions were considered. finally, it is presented that the NRB and HDRB systems are effective for the small equipment isolation and the damping of isolation systems can be affected to the seismic isolation effect.

• Proceedings of the Korean Institute of Building Construction Conference
• /
• 2017.11a
• /
• pp.183-184
• /
• 2017

Some of the information created during the design phase of an New NPP life cycle is useful only for the execution of the construction phase; however, much of the information greatly impacts the longer-term operational phase. To most make use of design and construction information produced by data based design system during the construction and operation phase, This research is identified controlled data and drawn design information of controlled equipment from documents generated during the life-cycle stages. This study aimed to analyze related documents to assure traceability of controlled equipment from design phase through O&M and then suggested DB(Data Base) based control method on technical information of major equipment throughout nuclear power plant lifecycle.

• Fire Science and Engineering
• /
• v.32 no.1
• /
• pp.57-65
• /
• 2018

In this study, a post-fire safe shutdown analysis (PFSSA) including multiple spurious operation (MSO) treatments for cables was conducted with an imaginary nuclear power plant for training using a deterministic fire analysis code. The imaginary nuclear power plant for the training consisted of a reactor containment building and an auxiliary building, including a total of 22 fire areas. The equipment including valves, pumps, emergency diesel generators, switch gears, motor control centers, and logic controllers were located in each fire area of the imaginary plant. It was assumed that each equipment is connected with two cables and that each cable passes through the fire areas along the cable trays. A database containing the information on the equipment and cables for the imaginary plant was constructed for the fire area analysis. The fire area analysis was performed for several assumed MSO scenarios, equipment logics, and cable logics. A mitigation measure using a three hour rated wrap was applied to the failed cables and cable trays after the fire area analysis.

• Journal of the Korean Society of Radiology
• /
• v.17 no.7
• /
• pp.1189-1195
• /
• 2023

In scenarios involving inspections and verifications of nuclear facilities, ensuring the proper functioning of on-site safeguards equipment is crucial. There have been precedents in Kazakhstan where equipment failed to operate properly due to extremly cold temperatures, and the year-round minimum temperature at North Korea's Punggye-ri nuclear test site is approximately minus 30 degrees Celsius. To ensure the proper functioning of equipment in extreme environments for on-site verification of nuclear activities on the Korean Peninsula, relevant research is necessary. This includes confirming the functionality of equipment used in inspections and verifications, as well as analyzing factors that may disrupt their normal operation. This study aims to conduct a risk analysis for the normal operation of equipment in extreme environments and develop criteria and procedures for environmental-based performance testing. To achieve this, we conducted a risk analysis based on IAEA safeguards, analyzed the utilization of equipment, and performed a risk analysis associated with transportation for on-site verification considering the environmental characteristics of the Korean Peninsula. Furthermore, we provided performance testing criteria and procedures. The research results can be utilized as reference material in the verification and monitoring processes of nuclear activities.

• Nuclear Engineering and Technology
• /
• v.37 no.2
• /
• pp.191-200
• /
• 2005

Shaking table tests of the seismic behavior of a steel frame structure model were performed. The purpose of these tests was to estimate the effects of a near-fault ground motion and a scenario earthquake based on a probabilistic seismic hazard analysis for nuclear power plant structures. Three representative kinds of earthquake ground motions were used for the input motions: the design earthquake ground motion for the Korean nuclear power plants, the scenario earthquakes for Korean nuclear power plant sites, and the near-fault earthquake record from the Chi-Chi earthquake. The probability-based scenario earthquakes were developed for the Korean nuclear power plant sites using the PSHA data. A 4-story steel frame structure was fabricated to perform the tests. Test results showed that the high frequency ground motions of the scenario earthquake did not damage the structure at the nuclear power plant site; however, the ground motions had a serious effect on the equipment installed on the high floors of the building. This shows that the design earthquake is not conservative enough to demonstrate the actual danger to safety related nuclear power plant equipment.

• Journal of the Korean Society of Systems Engineering
• /
• v.14 no.2
• /
• pp.73-82
• /
• 2018

In this work, a hardware based cryptographic module for the cyber security of nuclear power plant is developed using a system engineering approach. Nuclear power plants are isolated from the Internet, but as shown in the case of Iran, Man-in-the-middle attacks (MITM) could be a threat to the safety of the nuclear facilities. This FPGA-based module does not have an operating system and it provides protection as a firewall and mitigates the cyber threats. The encryption equipment consists of an encryption module, a decryption module, and interfaces for communication between modules and systems. The Advanced Encryption Standard (AES)-128, which is formally approved as top level by U.S. National Security Agency for cryptographic algorithms, is adopted. The development of the cyber security module is implemented in two main phases: reverse engineering and re-engineering. In the reverse engineering phase, the cyber security plan and system requirements are analyzed, and the AES algorithm is decomposed into functional units. In the re-engineering phase, we model the logical architecture using Vitech CORE9 software and simulate it with the Enhanced Functional Flow Block Diagram (EFFBD), which confirms the performance improvements of the hardware-based cryptographic module as compared to software based cryptography. Following this, the Hardware description language (HDL) code is developed and tested to verify the integrity of the code. Then, the developed code is implemented on the FPGA and connected to the personal computer through Recommended Standard (RS)-232 communication to perform validation of the developed component. For the future work, the developed FPGA based encryption equipment will be verified and validated in its expected operating environment by connecting it to the Advanced power reactor (APR)-1400 simulator.

• 제어로봇시스템학회:학술대회논문집
• /
• 2000.10a
• /
• pp.288-288
• /
• 2000

Reactor vessel is one of the most important equipment with regard to the safety of nuclear power plant. Thus nuclear regulation requires its periodical examination by certified inspection experts. Conventional reactor inspection machines are obsolete, hard to handle, and very expensive. To solve these problems we developed robotic reactor vessel inspection system which are small, easy to use for inspection, cost effective, and convenient in operation. This paper describes the main features of Main Control System which is one part of robotic inspection equipment we developed.