• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.18 no.2
• /
• pp.50-53
• /
• 2022

To investigate the validity of the finite element analysis program to assess structural integrity of a spent nuclear fuel transport cask subjected to extreme impact loads, structural integrity of the cask for the case of an aircraft engine collision is evaluated using three FE analysis programs: Autodyn, Speed and ABAQUS explicit version. As a result of all analyses, it is confirmed that no penetration occurred in the cask wall. Even though the different programs are used, it is identified that there are insignificant differences in the FE analysis variables such as von Mises effective stress and equivalent plastic strain among the programs.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.19 no.2
• /
• pp.69-74
• /
• 2023

The paper presents the results investigating the effect of aircraft engine impact location on the intended function evaluation results of spent nuclear fuel transport cask. As a result of the investigation, it is found that the structural integrity is maintained as the maximum accumulated equivalent plastic strain is below the acceptable criterion regardless of the collision location. It is identified that when the aircraft engine collided with the upper part of the transport cask without considering impact limiter the containment performance is weakened compared to when the aircraft engine collided with the central part.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.37 no.7
• /
• pp.875-881
• /
• 2013

The failure of a bolted joint owing to stress corrosion cracking (SCC) has been considered one of the most important structural integrity issues in a nuclear power plant. In this study, the failure possibility of bolting, which is used to support the steam generator of a pressurized water reactor, owing to SCC and brittle fracture was evaluated in accordance with guidelines proposed by the Electric Power Research Institute, which are called the Reference Flaw Factor method. For this evaluation, first, detailed finite element stress analyses were conducted to obtain the actual nominal stresses of bolting in which either service loads or bolt preloads were considered. Based on these nominal stresses, the structural integrity of bolting was addressed from the viewpoints of SCC and toughness. In addition, the accuracy of the EPRI Reference Flaw Factor for assessing bolting failure was investigated using finite element fracture mechanics analyses.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.1
• /
• pp.74-83
• /
• 2020

A number of so-called 'Type IV' cracking was reported to occur at the welded joints of the P91 steel or P92 steel reheat steam piping systems in Korean supercritical thermal power plants. The reheat steam piping systems are subjected to severe thermal and pressure loading conditions of coolant higher than 570℃ and 4MPa, respectively. In this study, piping analyses and design evaluations were conducted for the piping system of a specific thermal plant in Korea and suggestions were made how structural integrity could be improved so that type IV cracks at the welded joints could be prevented. Integrity evaluations were conducted as per ASME B31.1 code with implicit consideration of creep effects which was used in original design of the piping system and as per nuclear-grade RCC-MRx code with explicit consideration of creep effects. Comparisons were made between the evaluation results from the two design rules. Another approach with modification or reduction of the redundant supports in the piping systems was investigated as a tool to mitigate thermal stresses which should essentially contribute to prevention of Type IV cracking without major modification of the existing piping systems. In addition, a post weld heat treatment method and repair weld method which could improve integrity of the welded joint of P91 steel were investigated.

• Proceedings of the KIEE Conference
• /
• 2006.04b
• /
• pp.179-181
• /
• 2006

각 원자력 발전소에서는 정밀성, 안전성을 확인하는 검사의 중요성을 인식하여 LPMS(Loose Part Monitoring System)을 사용하여 사고 징후를 조기에 감지하여 이에 대한 예방조치를 가능케 함으로써 설계기준 사고 발생을 사전에 방지할 수 있게 한다. 또한 이 기술은 신호 측정 및 분석 등의 기반기술 개발을 통하여 건전성 감시 기술의 신뢰성을 향상 시키고 있다. LPMS(Loose Part Monitoring System)기술은 재료, 기기, 구조물 등의 성질과 내부조직을 변화시키거나 파괴하지 않고, 배관내부에 흐르는 금속 파편들을 찾아내어 정밀성, 안전성, 신뢰성을 확보하기 위하여 검사기술이 적용되고 있다. 그러나 이 방법은 배관내의 이물질의 충격이 발생해야 감지가 가능하고, 이물질의 모양이나 사이즈를 확인하기에는 어려움이 있다. 따라서 본 논문에서는 배관외부에서 자기장을 인가하여, 배관내의 이물질에 변화하는 자기장을 홀센서로 측정하여 기존의 LPMS 방식을 보완하는 시스템을 개발하기 위해, 배관에 필요한 자기장 발생장치를 설계하고, 이물질을 검출하기 위한 검출 감도향상에 대해 연구하였다.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.13 no.2
• /
• pp.1-18
• /
• 2017

Metal fatigue is an important aging mechanism that material characteristics can be deteriorated when even a small load is applied repeatedly. An accurate fatigue evaluation is very important for component structural integrity and reliability. In the design stage of a nuclear power plant, the fatigue evaluations of the Class 1 components have to be performed. However, operating experience shows that the design evaluation can be very conservative due to conservatism in the transient severity and number of occurrence. Therefore, the fatigue monitoring system has been considered as a practical mean to ensure safe operation of the nuclear power plants. The fatigue monitoring system can quantify accumulated fatigue damage up to date for various plant conditions. The purpose of this paper is to describe the fatigue monitoring procedure and to introduce the fatigue monitoring program developed by the authors. The feasibility of the fatigue monitoring program is demonstrated by comparing with the actual operating data and finite element analysis results.

• Journal of the Korea Society for Simulation
• /
• v.19 no.2
• /
• pp.147-155
• /
• 2010

It often occur to nature disaster that like earthquake, typhoon, etc. around KOREA. A Haiti and Chile also metropolitan area of KOREA occur earthquake. in result, People consider nature disaster. Structures of present age are easily affected by nature disaster. So we are important that warn of dangerous situation as soon as possible. On this study, I introduce Integrated Health Monitoring System for Infra-structure. I develop Structure Health Monitoring System on web-site. Administrator always monitor structure on real-time using internet network. As Administrator using mobile device like PDA, Administrator always monitor structure. As using this system, Damage of nature disaster is minimized and is prevented post damage.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.20 no.8
• /
• pp.23-28
• /
• 2019

The structural integrity of a surface metal coating was evaluated through numerical results to improve the efficiency and reduce the damage caused by cavitation in ships and marine plants. The goal was to ensure structural strength and performance, even if the thickness of the wing is reduced to reduce the weight of the material and surface coating. Analytical methods were used for four models: a non-coating model, one with the same thickness after coating, one with a thickness reduction of 3% after coating, and one with thickness reduction of 5% after coating. With a thickness reduction of 5% after coating, the stress was increased to 12%, and the safety factor was 0.99%, so the structural integrity was insufficient. However, a better material or a thicker coating could allow a sufficient safety factor to be secured. The structural integrity was improved by the coating, and even when the weight was reduced up to 5%, the structural integrity could be sufficiently secured due to the coating effect.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.12 no.1
• /
• pp.70-77
• /
• 2016

In this paper, the structural design and integrity evaluations for a reactor vessel of PGSFR sodium-cooled fast reactor(150MWe) are carried out in compliance with ASME BPV III, Division 5 Subsection HB. The reactor vessel is designed with a direct contact of primary sodium coolant to its inner surface and has a double vessel concept enclosing by containment vessel. To assure the structural integrity for 60 years design lifetime and elevated operating temperature of $545^{\circ}C$, which can invoke creep and creep-fatigue damage, the structural integrity evaluations are carried out in compliance with the ASME code rules. The design loads considered in this evaluations are primary loads and operation thermal cycling loads of normal heat-up and cool-down. From the evaluations, the PGSFR reactor vessel satisfies the ASME code limits but it was found that there is a little design margin of creep damage for inner surface at the region of cold pool free surface.

• The Journal of the Korea institute of electronic communication sciences
• /
• v.13 no.4
• /
• pp.751-756
• /
• 2018

In smart factories and industrial IoT, all facilities in a factory are monitored over the Internet, thereby facility can reduce the downtime and increase the availiability by preventive maintenance before it breaks down. The abnormal conditions of the major facilities in the plant are caused by abnormal temperature rise, vibration, and variations in noise. Consequently, it is critical to develop a very small smart device that is easily installed in a small space to enable real-time monitoring of the vibration status of the facility. In this study, smart devices were developed for smart factory fault prediction and robustness management using ultra small micro-controllers with WiFi capabilities and MEMS acceleration sensors.