• 품질경영학회지
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• 제18권2호
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• pp.69-80
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• 1990

This paper highlights the economic effects of nuclear power plants standardization in Korea. The major effects of nuclear power plants standardization appear in the reduction of time-related costs. By using this major economic effects of standardization, an optimal plant mix of electric power until the year 2005 is suggested by means of WASP computer model. And the effects between the standardized case and the non-standardized case is compared.

• 대한인간공학회지
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• 제33권1호
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• pp.49-58
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• 2014

Objective: The aim of this study is to enhance the efficiency of education and training through application and management of 'Transfer of Training' in nuclear power plants. Background: Despite the sophistication and standardization of job-related skills and techniques of workers, accidents/incidents keep taking place due to human errors and unsafe actions and behaviors, which translates into the necessity to review and examine the effectiveness and influence of education and training on the workers of nuclear power plants. Method/Results: This study drew the factors of 'Transfer of Training' through a review on the preceding studies and document research. In addition, through expert examination, this study explored the expected effects and possibility of application when managing the influencing factors of 'Transfer of Training' in nuclear power plants. And lastly, management priority order for nuclear power plants was drawn through an AHP analysis. Conclusion: Among the 'Transfer of Training' factors, the training design factor was the most important. In addition, the design of the training and transfer and goal setting showed a high degree of importance among the influencing factors. Application: The management of 'Transfer of Training' in nuclear power plants enhances the capability of workers and improves the operational integrity of nuclear power plants.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.235-236
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• 2017

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.605-610
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• 2005

A method to estimate the relative sensitivity of a self-powered rhodium detector for an upcoming cycle is developed by combining the rhodium depletion data from a nuclear design with the site measurement data. This method can be used both by nuclear power plant designers and by site staffs of Korean standard nuclear power plants for determining which rhodium detectors should be replaced during overhauls.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.173-173
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• 2003

• Nuclear Engineering and Technology
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• 제24권2호
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• pp.183-192
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• 1992

한국 표준 원전의 부하추종운전 능력 향상을 위하여 노심 축 방향 출력분포 제어에 기존의 regulating bank와는 별개의 heavy bank를 사용하는 노심 제어 기법인 Mode K를 제시하였다 이 경우 heavy bank는 제어할 수 있는 축방향 출력 변화의 범위가 크며, 또한 bank의 움직임과 축 방향 출력 변화가 단조적 관계를 항상 유지할 수 있으므로 운전을 용이하게 할 뿐 아니라 자동제어가 가능하도록 한다. Mode K기법을 이용한 노심 제어 자동화는 부하추종운전에 대한 운전원의 부담을 경감시킬 수 있으며, 특히 주파수 제어 운전등 미리 예측되지 않은 부하변동에 응동할 수 있는 능력을 갖게 한다. 표준 원전의 참조 발전소인 영광 3,4호기의 운전을 모의 계산함으로써 Mode K 의 설계 개념을 평가하였는데 그 결과를 통하여 한국 표준 원전의 실제적 부하추종운전을 위하여 Mode K 기법을 적절하게 사용할 수 있음을 입증하였다.

• Journal of Radiation Protection and Research
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• 제28권3호
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• pp.239-245
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• 2003

Every nuclear power plant measured concentrations of tritium in groundwater and surface water around the plants periodically. It was not easy to predict the tritium concentration only with these measurement data in case of various release scenarios. KAERI developed a new approach to find the relationship between the tritium release rate and tritium concentration in the environment. The approach was based upon a dynamic compartment model. In this paper the dynamic compartment model was modified to predict the tritium behavior more accurately. The mechanisms considered for the transfer of tritium between the compartments were evaporation, groundwater flow, infiltration, runoff, and hydrodynamic dispersion. Time dependent source terms of the compartment model were introduced to refine the release scenarios. Also, transfer coefficients between the compartments were obtained using realistic geographical data. In order to illustrate the model various release scenarios were developed, and the change of tritium concentration in groundwater and surface water around the nuclear power plants was estimated.

• 한국신뢰성학회:학술대회논문집
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• 한국신뢰성학회 2002년도 정기학술대회
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• pp.387-401
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• 2002

The Digital Engineered Safety Feature Actuation System (DESFAS) of nuclear power plants actuates safety systems to mitigate severe accidents occurred in nuclear power plants. The reliability of the system should be evaluated in order to meet the reliability criteria of nuclear power plants. In this work, we have calculated and evaluated the lifetime of DESFAS by using Reliability Block Diagram (RBD) and failure rates of digital control components. Surveillance test is assumed in the evaluation. The result shows that the digital control component can be used in DESFAS system.

• Nuclear Engineering and Technology
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• 제37권3호
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• pp.259-264
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• 2005

This paper presents a vital area identification (VAI) method based on the current fault tree analysis (FTA) and probabilistic safety assessment (PSA) techniques for the physical protection of nuclear power plants. A structured framework of a top event prevention set analysis (TEPA) application to the VAI of nuclear power plants is also delineated. One of the important processes for physical protection in a nuclear power plant is VAI that is a process for identifying areas containing nuclear materials, structures, systems or components (SSCs) to be protected from sabotage, which could directly or indirectly lead to core damage and unacceptable radiological consequences. A software VIP (Vital area Identification Package based on the PSA method) is being developed by KAERI for the VAI of nuclear power plants. Furthermore, the KAERI fault tree solver FTREX (Fault Tree Reliability Evaluation eXpert) is specialized for the VIP to generate the candidates of the vital areas. FTREX can generate numerous MCSs for a huge fault tree with the lowest truncation limit and all possible prevention sets.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.919-928
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• 2012

The applications of computers and communication system and network technologies in nuclear power plants have expanded recently. This application of digital technologies to the instrumentation and control systems of nuclear power plants brings with it the cyber security concerns similar to other critical infrastructures. Cyber security risk assessments for digital instrumentation and control systems have become more crucial in the development of new systems and in the operation of existing systems. Although the instrumentation and control systems of nuclear power plants are similar to industrial control systems, the former have specifications that differ from the latter in terms of architecture and function, in order to satisfy nuclear safety requirements, which need different methods for the application of cyber security risk assessment. In this paper, the characteristics of nuclear power plant instrumentation and control systems are described, and the considerations needed when conducting cyber security risk assessments in accordance with the lifecycle process of instrumentation and control systems are discussed. For cyber security risk assessments of instrumentation and control systems, the activities and considerations necessary for assessments during the system design phase or component design and equipment supply phase are presented in the following 6 steps: 1) System Identification and Cyber Security Modeling, 2) Asset and Impact Analysis, 3) Threat Analysis, 4) Vulnerability Analysis, 5) Security Control Design, and 6) Penetration test. The results from an application of the method to a digital reactor protection system are described.