• Nuclear Engineering and Technology
• /
• 제19권2호
• /
• pp.99-106
• /
• 1987

가상적인 1,000MWe의 석탄화력발전과 원자력 발전소로부터 배출되는 방사성물질에 의한 피폭 영향을 상호 비교 연구하였다. 본 논문에서는 정상가동중에 배출되는 기체상 방사성물질에 국한하였으며 석탄화력발전소에 대한 방사선원은 국내자료가 부족하여 외국자료에 근거했고, 원자력발전소에 대해서는 표준발전소에 대해 계산된 방사선원을 사용하였다. 고리 기상탑의 1년 기상자료를 이용하여 Gaussian모델에 의해 방사성물질의 대기확산을 평가했으며, 개인 피폭선량은 대기확산인자가 최대인 지점의 성인에 대해 계산하였다. 방사선피폭선량은 석탄화력발전소보다 원자력발전소의 경우가 약간 컸으며 석탄화력의 경우는 원자력발전소와 달리 피폭선량의 73.5%가 오염된 엽채류의 섭취에 따른 것이었다.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.502-507
• /
• 1997

As a maintenance optimization project to improve the safety of Wolsong NPP (Nuclear Power Plant), reliability of diesel generators are estimated based on the operating experience, and improvement options are suggested. A reliability measure is suggested for the estimation of reliability for standby safety systems to reflect availability. It is assessed that the reliability of diesel generators can be much improved if the suggested improvement options are implemented.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.631-636
• /
• 1996

This paper presents a Reliability Centered Maintenance (RCM) framework for implementation on safety system of nuclear power plant (NPP). RCM is a systematic methodology to optimize the surveillance and maintenance tasks for critical components which provides efficiently and effectively reliability of system and safety of plant. Maintenance of the safety systems is essential for its safe and reliable operation. Reliability Centered Maintenance at NPP is the program which assure that plant system remains within original design criteria and that is not adversely affected during the plant life time. Aim of this paper is to provide the RCM framework to implement it on safety systems. RCM framework is described in four major steps.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2004년도 추계학술발표회 발표논문집
• /
• pp.605-606
• /
• 2004

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1995년도 추계학술발표회논문집(1)
• /
• pp.423-428
• /
• 1995

Simulation models have been developed to predict the overall behavior of the CANDU plant systems during normal operational transients. For real time simulation purpose, simplified thermal hydraulic models are applied with appropriate system control logics, which include primary heat transport system solver with its component models and secondary side system models. The secondary side models are mainly used to provide boundary conditions for primary system calculation and to accomodate plant power control logics. Also, for the effective use of simulation package, hardware oriented basic simulation network has been established with appropriate graphic display system. Through validation with typical plant power maneuvering cases using proven plant performance analysis computer code, the present simulation package shows reasonable capability in the prediction of the dynamic behavior of plant variables during operational transients of CANDU plant, which means that this simulation tool can be utilized as a basic framework for full scope simulation network through further improvements.

• 대한인간공학회지
• /
• 제25권3호
• /
• pp.85-96
• /
• 2006

Main control room(MCR) man-machine interface(MMI) design of advanced nuclear power plants(NPPs) such as APR(advanced power reactor)-1400 can be validated through performance-based tests to determine whether it acceptably supports safe operation of the plant. In this work, plant performance, personnel task, situation awareness, workload, teamwork, and anthropometric/physiological factor are considered as factors for the human performance evaluation. For development of measures in each of the factors, techniques generally used in various industries and empirically proven to be useful are adopted as main measures and some helpful techniques are developed in order to complement the main measures. Also the development of the measures is addressed based on the theoretical background. Finally we discuss the way in which the measures can be effectively integrated and then HUPESS(HUman Performance Evaluation Support System) which is in development based on the integrated way is briefly introduced.

• Nuclear Engineering and Technology
• /
• 제54권8호
• /
• pp.2840-2851
• /
• 2022

Electric boilers (EBs) are the backup steam source for the auxiliary steam system of high-temperature gas-cooled reactor nuclear power plants. When the plant is in normal operations, the EB is always in hot standby status. However, the current hot standby operation strategy has problems of slow response, high power consumption, and long operation time. To solve these problems, this study focuses on the optimization of hot standby operations for the EB system. First, mathematical models of an electrode immersion EB and its accompanying deaerator were established. Then, a control simulation platform of the EB system was developed in MATLAB/Simulink implementing the established mathematical models and corresponding control systems. Finally, two optimization strategies for the EB hot standby operation were proposed, followed by dynamic simulations of the EB system transient from hot standby to normal operations. The results indicate that the proposed optimization strategies can significantly speed up the transient response of the EB system from hot standby to normal operations and reduce the power consumption in hot standby operations, improving the dynamic performance and economy of the system.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2003년도 춘계학술대회
• /
• pp.610-616
• /
• 2003

Recently, the common concern of nuclear power industry in the development of technology mitigating and preventing the aging of steam generator tubes prevails, because the trends of steam generator flaws at Uljin unit #1,2 and KSNP(Korea Standard Nuclear Power Plant) impose a burden on the operation of nuclear power plant. While the regulatory agency is demanding the establishment of the advanced general performance maintenance system, the steam generator management program adapting advanced technology is being developed which may comply with EPRI PWR SG Guidelines based on NEI 97-06 ‘ General Guidelines including all the maintenance aspects consist of the tube integrity assessment criteria, repair limit, allowable leakage level, water chemistry will be composed in order to obtain the approval of regulatory agency and be applied to Nuclear power plant early 2005. This presentation is to introduce maintenance state including SG tube degradation and main contents of advanced SG management program being developed, and futhermore update present and future plan, and estimate the alternation after the completion.

• 한국압력기기공학회 논문집
• /
• 제8권1호
• /
• pp.25-32
• /
• 2012

Korea Hydro & Nuclear Power Co. (KHNP) has developed and implemented the maintenance effectiveness monitoring (MR) programs for the operating nuclear power plants. The MR program is developed by reflecting design characteristics of the operating nuclear power plants to monitor the plant performance for improving the safety and reliability. Recently, KHNP has built a new nuclear power plant, and developed the MR program to establish the advanced maintenance system by reflecting unique design characteristics based on the OPR1000 standard model. So, the MR program developed in this study has another characteristics in comparison with the OPR1000 standard model, and we will verify the suitability of the MR program through evaluating initial performance of the plant. The safety and reliability of the new plant will be improved by developing and implementing the MR program.

• 시스템엔지니어링학술지
• /
• 제12권1호
• /
• pp.89-103
• /
• 2016

Due to safety of the plant, it became very clear the importance of study occurrence reactor coolant system (RCS) issues specially the primary water stress corrosion cracking (PWSCC). The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. Robotic devices have been used for internal inspection, maintenance and performing remote welding and inspection in high-radiation areas. In this paper, PWSCC overview and inlay and over lay welding methodology introduced, concept of robotic device that can be inserted into the piping via Steam Generator (SG) main way to access to primary piping of pressurized water reactor (PWR) is developed based on SE methodology. A 3D model of the inspection system was developed along with the APR1400 (Advanced Power Reactor)reactor coolant systems (RCS) and internals with virtual 3D simulation of the operation for visualization to prove the validity of the concept.