• Nuclear Engineering and Technology
• /
• 제55권6호
• /
• pp.2215-2221
• /
• 2023

In order to further meet the requirements of weight, volume, and dose minimization for new nuclear energy devices, the bare-bones multi-objective particle swarm optimization algorithm is used to automatically and iteratively optimize the design parameters of radiation shielding system material, thickness, and structure. The radiation shielding optimization program based on the bare-bones particle swarm optimization algorithm is developed and coupled into the reactor radiation shielding multi-objective intelligent optimization platform, and the code is verified by using the Savannah benchmark model. The material type and thickness of Savannah model were optimized by using the BBMOPSO algorithm to call the dose calculation code, the integrated optimized data showed that the weight decreased by 78.77%, the volume decreased by 23.10% and the dose rate decreased by 72.41% compared with the initial solution. The results show that the method can get the best radiation shielding solution that meets a lot of different goals. This shows that the method is both effective and feasible, and it makes up for the lack of manual optimization.

• 한국안전학회지
• /
• 제26권4호
• /
• pp.7-13
• /
• 2011

A study on reliability estimation of sequential-ordered multiple failure modes, which are sequentially ordered between failure modes in a considering system, was performed. Especially, an approach to estimate the probabilities of failure modes has been proposed under an assumption that failure modes are mutually exclusive and sequentially ordered by only a critical variable. A feasibility of the proposed approach were studied by a practical example, which is a reliability estimation of passive safety systems for a probabilistic safety assessment(PSA) of a very high temperature reactor(VHTR) that is under development as a future nuclear system with enhanced safety features. It is difficult to define a robust failure state of this nuclear system because of its enhanced radiation release characteristics, so the new approach is a useful concept to estimate not only its safety but also a PSA. A feasibility study applied two failure modes(e.g., small and large release of radioactive materials) with considering the integrated behavior of this nuclear system. It is expected that the multiple release states for a practical estimation can be easily extended to the aforementioned example. It was found out that the proposed approach was a useful technique to cover the unfavorable features of this nuclear system as to performing a VHTR PSA.

• 수산경영론집
• /
• 제50권2호
• /
• pp.23-40
• /
• 2019

The main purpose of this study is to analyze and examine the problems of the law systems of the safety and maintenance of nuclear facilities and to propose the improvements with respect to the related problems especialy focused on safety management of aquatic products. Therefore, the results of the paper would be helpful to build an effective management law system of safety and maintenance of nuclear facilities and fisheries products. The research methods are longitudinal and horizontal studies. This study compares domestic policies with foreign policies of nuclear plants and aquatic products. Using the above methods, examining the current system of nuclear-related laws and regulations, we have found that there exist 13 Acts including "Nuclear Safety Act", etc. Safety laws related on nuclear facilities have seven Acts including "Nuclear Safety Act", "the Act on Physical Protection and Radiological Emergency", "Radioactive waste control Act", "Act on Protective Action Guidelines against Radiation in the Natural Environment", "Special Act on Assistance to the locations of facilities for disposal low and intermediate level radioactive waste", "Korea Institute of Nuclear Safety Act". "Act on Establishment and Operation of the Nuclear Safety and Security Commission". The seven laws are composed of 119 legislations. They have 112 lower statute of eight Presidential Decrees, six Primeministrial Decrees and Ministrial Decrees, 92 administrative rules (orders), 6 legislations of local self-government aself-governing body. The concluded proposals of this paper are as follows. Firstly, we propose that the relationship between the special law and general law should be re-established. Secondly, the terms with respect to law system of safety and maintenance of nuclear plants should be redefined and specified. Thirdly, it is advisable to re-examine and re-establish the Law System for Safety and Maintenance of Nuclear Facilities. and environmental rights like the French Nuclear Safety Legislation. Lastly, inadequate legislation on the aquatic pollution damage should be re-established. It is necessary to ensure sufficient transparency as well as environmental considerations in the policy decisions of the Korean government and legislation of the National Assembly. It is necessary to further study the possibilities of accepting the implications of the French legal system as a legal system in Korea. In conclusion, the safety management of nuclear facilities is not only focused on the secondary industry and the tertiary industry centering on power generation and supply, but also on the primary industry, which is the food of the people. It is necessary to prevent damage to be foreseen. Therefore, it is judged that there should be no harm to the people caused by contaminated marine products even if the "Food Safety Law for Prevention of Radiation Pollution Damage" is enacted.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 추계학술발표회
• /
• pp.586-587
• /
• 2005

• 대한인간공학회지
• /
• 제33권2호
• /
• pp.77-86
• /
• 2014

Objective: The purpose of this study is to grasp the fundamental structure of incident occurrence in nuclear organizations based on system thinking, and analyze how various causes are interrelated in terms of the causal loop diagram. Background: The recent domestic and overseas nuclear power plant-related incidents and accidents are directly or indirectly associated with safety culture, and thus effective plans for the improvement of safety culture are being called for. While the safety of a nuclear power plant is highly dependent upon technology and equipment, the utilization, maintenance and inspection of the technology and equipment are conducted by workers of the nuclear power plant. Method: Methodology of system thinking perspective using causal loop analysis. Results: As a result of the analysis, first, it turned out that the fundamental cause of incident occurrence in nuclear organizations is time constraint. Second, if a workload of workers increases, their adherence to regulations and procedures comes to be reduced due to time constraint. Third, it is needed, through organizational learning education, to increase actions made from thoughts considering safety as the utmost priority in advance. Fourth, it is necessary to improve professionalism by enhancing educational programs for new workers, and to develop various scenarios with which they can cope with certain situations. Application: This paper provides a base for system dynamics simulation model for future study.

• Nuclear Engineering and Technology
• /
• 제48권5호
• /
• pp.1174-1183
• /
• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

• Nuclear Engineering and Technology
• /
• 제52권1호
• /
• pp.78-86
• /
• 2020

The operation and monitoring (OM) interface is the digital medium between nuclear power system and operators. The cognitive load of OM interface has an important effect on the operation errors made by operator during OM task between operator and computer. The cognitive load model of OM interface is constructed for analysing the composition and influencing factors of OM interface cognitive load. And to study the coping strategies and methods for cognitive load of nuclear power system. An experiment method based on eye movement is proposed to measure the cognitive load of OM interface. Experiment case is carried out with 20 subjects and typical OM interface of a nuclear power system simulator. The OM interface is optimized based on the experiment results. And the results comparison between the original OM interface and the optimized OM interface shows that the cognitive load model and proposed method is valuable contributions in reducing the cognitive load and improving the interaction efficiency of OM tasks.

• Nuclear Engineering and Technology
• /
• 제54권12호
• /
• pp.4560-4570
• /
• 2022

Dynamic fault tree (DFT) and its related research methods have received extensive attention in safety analysis and reliability engineering. DFT can perform reliability modelling for systems with sequential correlation, resource sharing, and cold and hot spare parts. A technical modelling method of DFT is proposed for modelling ship collision accidents and loss-of-coolant accidents (LOCAs). Qualitative and quantitative analyses of DFT were carried out using the cutting sequence (CS)/extended cutting sequence (ECS) method. The results show nine types of dynamic fault failure modes in ship collision accidents, describing the fault propagation process of a dynamic system and reflect the dynamic changes of the entire accident system. The probability of a ship collision accident is 2.378 × 10^-9 by using CS. This failure mode cannot be expressed by a combination of basic events within the same event frame after an LOCA occurs in a marine nuclear reactor because the system contains warm spare parts. Therefore, the probability of losing reactor control was calculated as 8.125 × 10^-6 using the ECS. Compared with CS, ECS is more efficient considering expression and processing capabilities, and has a significant advantage considering cost.

• 산업경영시스템학회지
• /
• 제33권4호
• /
• pp.114-121
• /
• 2010

Traditionally Nuclear Research and Development (R&D) result has been big influence on other industries and societies and it requires large scale investments and study period. So it is essential to apply Quality Assurance (QA) for systematic R&D management. This paper investigates QA System for U.S. Nuclear R&D and reviews QA elements. Based on this investigation, we applied QA requirements to Nuclear Hydrogen R&D project, and the scope of application be enlarged as R&D stage in progress. We also present QA system improvement way through consideration for Nuclear Hydrogen Project's QA application. As the need for QA in R&D is expected to increase in the future, it is necessary to prepare guidelines for R&D QA.

• Nuclear Engineering and Technology
• /
• 제41권1호
• /
• pp.71-78
• /
• 2009

One of main concepts involved in regional small nuclear reactors is unmanned remote control. Internet-based virtual private networks provide environments for the remote monitoring and control of geographically-dispersed systems, and with the advances in communication technologies, the potential of networks for real time control and automation becomes enormous. However, networked control has some problems. The most critical is delay in signal transmission, which degrades system stability and performance. Therefore, a networked control system should be designed to account for delay. This paper proposes some design approaches for a delay-tolerant system that can guarantee predetermined stability margins and performance. To accomplish this, the reactor plant is modeled with consideration of uncertainties. With this model, three kinds of controllers are developed using different methods. The designed systems are compared with respect to stability and performance, and a second-order controller designed using the table lookup method was found to give the most satisfactory results.