• Structural Engineering and Mechanics
• /
• v.58 no.1
• /
• pp.39-58
• /
• 2016

Since the publication of ACI 318-02, the concrete capacity design (CCD) method has been used to determine the resistance of unreinforced concrete anchors. The regulation of steel-reinforced anchors was proposed in ACI 318-08. Until ACI 318-08, the shear resistance of concrete breakout for an unreinforced anchor during an earthquake was reduced to 75% of the static shear strength, but this reduction has been eliminated since ACI 318-11. In addition, the resistance of a hairpin-reinforced anchor was calculated using only the strength of the steel, and a regulation on the dynamic strength was not given for reinforced anchors. In this study, shaking table tests were performed to evaluate the dynamic shear strength of unreinforced and hairpin-reinforced cast-in-place (CIP) anchors during earthquakes. The anchors used in this study were 30 mm in diameter, with edge distances of 150 mm and embedment depths of 240 mm. The diameter of the hairpin steel was 10 mm. Shaking table tests were carried out on two specimens using the artificial earthquake, based on the United States Nuclear Regulatory Commission (US NRC)'s Regulatory Guide 1.60, and the Northridge earthquake. The experimental results were compared to the current ACI 318 and ETAG 001 design codes.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2012.04a
• /
• pp.734-738
• /
• 2012

A comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals is established in accordance with the United States Nuclear Regulatory Commission Regulatory Guide 1.20 Revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results at the inner barrel assembly top plate in the reactor. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at Inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.22 no.5
• /
• pp.474-479
• /
• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) reactor vessel internals is established in accordance with the united states nuclear regulatory commission regulatory guide 1.20 revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results of the inner barrel assembly top plate in APR1400. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for APR1400 reactor vessel internals.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2014.10a
• /
• pp.99-101
• /
• 2014

A Comprehensive Vibration Assessment Program (CVAP) for steam generator internals (SGI) of Advanced Power Reactor 1400 (APR1400) is being performed in accordance with the United States Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.20 (RG 1.20) revision 3. This paper studies the vibration characteristics of moisture separator assembly as part of the vibration and stress analysis program for APR1400 SGI CVAP. The natural frequencies, mode shapes, and structural behavior of moisture separator assembly were investigated through modal analysis using finite element method and experimental measurement. Since the moisture separator consists of several items with complicated shape, an idealized shell model was used in the finite element analysis. Group of local modes caused by moisture separators and significant modes of shroud and separator support plate were identified. The results of this paper are to be utilized in the structural response analysis of moisture separator assembly.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
• /
• 2022.05a
• /
• pp.187-188
• /
• 2022

As information processing technology develops with the trend of the times, the possibility of cyber threats to nuclear facilities is increasing. In the 2000s, there was a growing perception that cyberattacks on nuclear facilities were needed, and in fact, a cybersecurity regulatory system for nuclear power plants began to be established to prepare for cyberattacks. In Korea, in order to prepare for cyber threats, in 2013 and 2014, the Act on Protection and Radiation Disaster Prevention, Enforcement Decree, and Enforcement Rules of Nuclear Facilities, etc., and notices related to the Radioactive Disaster Prevention Act were revised. In 2015, domestic nuclear operators prepared information system security regulations for each facility in accordance with the revised laws and received approval from the Nuclear Safety Commission for implementation of information system security regulations divided into seven stages. In 2019, a special inspection for step-by-step implementation was completed, and since 2019, the cybersecurity system of operators has been continuously inspected through regular inspections. In this paper, we present some measures to build improved training to suit the steadily revised inspection of the nuclear facility cybersecurity system to counter cyber threats to the ever-evolving nuclear facilities.

• Nuclear Engineering and Technology
• /
• v.55 no.1
• /
• pp.1-11
• /
• 2023

To protect people and the environment from environmental radiation, the Act on Protective Action Guidelines against Radiation in the Natural Environment was formulated in Korea in 2011. This law regulates matters related to radiation safety that can be encountered in life. In accordance with this law, radiation monitoring equipment is operated at major airports and ports across the country, ensuring radiation monitoring of imported cargo. Currently, six ministries conduct radiation monitoring of imported cargo: the Nuclear Safety and Security Commission; the Korea Customs Service; the Ministry of Food and Drug Safety; the Ministry of Environment; the Ministry of Agriculture, Food and Rural Affairs; and the Korea Forest Service. Each ministry designates the relevant cargo items for radiation monitoring. The objective of this study was to comprehensively review the Korean radiation monitoring system for imported cargo and identify the areas and scopes of improvement. This paper also proposes a new law and an integrated supervision plan, which involves establishing a dedicated department to enhance the efficiency and professionalism of the national radiation monitoring system for imported cargo. The review will contribute to the development of a more sophisticated national radiation monitoring system for imported cargo.

• Proceedings of the Korean Information Science Society Conference
• /
• 2006.06c
• /
• pp.220-222
• /
• 2006

본 논문은 안전등급 제어기기(Safety-Grade Programmable Logic Controller)에서 사용하는 프로세서모듈 운영체제에 대한 컴포넌트 시험에 대해 기술한다. 디지털 소프트웨어에 대한 NRC(Nuclear Regulatory Commission)의 지침에 따라 운영체제는 소프트웨어 생명주기에 따라 개발되고 있으며 요구사항과 설계명세, 그리고 구현코드를 가지고 다양한 시험을 수행하고 있다. 컴포넌트 시험은 구현된 코드가 테스트 커버리지를 만족하는 지 파악하는 시험이다. 이를 위해 설계명세서를 참조하여 시험대상을 구분하고 각각의 시험대상에 대한 시험항목을 세분화한 이후 시험방법과 절차, 그리고 시험환경을 구축한 후 컴포넌트 시험을 수행한다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.1
• /
• pp.95-106
• /
• 2019

The decommissioning of nuclear power plants should be prepared by quantitative and qualitative risk assessment. Radiological and non-radiological hazards arising during decommissioning activities must be assessed to ensure the safety of decommissioning workers and the public. Decommissioning experiences by U.S. operators have mainly focused on deterministic risk assessment, which is standardized by the U.S. Nuclear Regulatory commission (NRC) and focuses only on the consequences of risk. However, the International Atomic Energy Agency (IAEA) has suggested an alternative to the deterministic approach, called the risk matrix technique. The risk matrix technique considers both the consequence and likelihood of risk. In this study, decommissioning stages, processes, and activities are organized under a work breakdown structure. Potential accidents in the decommissioning process of NPPs are analyzed using the composite risk matrix to assess both radiological and non-radiological hazards. The levels of risk for all potential accidents considered by U.S. NPP operators who have performed decommissioning were estimated based on their consequences and likelihood of events.

• Journal of Radiation Protection and Research
• /
• v.41 no.3
• /
• pp.260-267
• /
• 2016

Background: This study analyzes the differences in the annual averaged atmospheric dispersion factor and ground deposition factor produced using two classification methods of atmospheric stability, which are based on a vertical temperature difference and the standard deviation of horizontal wind direction fluctuation. Materials and Methods: Daedeok and Wolsong nuclear sites were chosen for an assessment, and the meteorological data at 10 m were applied to the evaluation of atmospheric stability. The XOQDOQ software program was used to calculate atmospheric dispersion factors and ground deposition factors. The calculated distances were chosen at 400 m, 800 m, 1,200 m, 1,600 m, 2,400 m, and 3,200 m away from the radioactive material release points. Results and Discussion: All of the atmospheric dispersion factors generated using the atmospheric stability based on the vertical temperature difference were shown to be higher than those from the standard deviation of horizontal wind direction fluctuation. On the other hand, the ground deposition factors were shown to be same regardless of the classification method, as they were based on the graph obtained from empirical data presented in the Nuclear Regulatory Commission's Regulatory Guide 1.111, which is unrelated to the atmospheric stability for the ground level release. Conclusion: These results are based on the meteorological data collected over the course of one year at the specified sites; however, the classification method of atmospheric stability using the vertical temperature difference is expected to be more conservative.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.6 no.4
• /
• pp.357-368
• /
• 2008

On July 31, 2008, the Government issued the construction and operation permit for the first low and intermediate level radioactive waste disposal facility in the Republic of Korea. In this paper, the fundamental regulatory framework, regulatory requirements and technical standards of the disposal facility are introduced, and the phased review process adopted for evaluation of the safety of the facility is briefly described. The Atomic Energy Act sets forth a stepwise regulatory framework for the whole life-cycle of the disposal facility such as siting, design, construction, operation, closure and institutional control. More detailed regulatory requirements and technical standards are stipulated in the subsequent regulations of the Atomic Energy Act and a series of Notices issued by the Ministry of Eduction, Science and Technology. The Korea Institute of Nuclear Safety, as entrusted by the Ministry under the Atomic Energy Act, conducted safety review on the disposal facility, and evaluated the compliance with relevant criteria in all technical elements(i.e. siting and structural safety, radiological environmental impact, operational safety, systems and components, quality assurance, and total systematic performance assessment, etc.). The overall safety review process can be phased into inception phase, initial review phase, main review phase and completion phase. The review results were reported to and deliberated by the five Sub-committees of the Special Committee on Nuclear Safety, and then reported to the Ministry. The Ministry issued the construction and operation permit of the disposal facility through the deliberation of the review results by the Nuclear Safety Commission. Hereafter, the safety of the repository will be reassured by a series of subsequent regulatory inspections and reviews under the Atomic Energy Act. In addition, the licensee's continuous implementation of the "Safety Promotion Plan" may also enhance the long-term safety of the repository and contribute to build-up the confidence of the safety case.