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

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

• 대한전기학회:학술대회논문집
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• 대한전기학회 1995년도 추계학술대회 논문집 학회본부
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• pp.172-174
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• 1995

About a half of Electric power is generated by nuclear power plants in korea. So, the stable operation of nuclear power plant is very important for suppling the essential national electric power. A S/G(Steam Generator) level control is the most difficult system in PWR(Pressurized Water Reactor) nuclear power plant. Because of the non-linear and the non-nominal response of S/G level control, it Is very difficult to control the level by automatic mode or manual mode. The goal of this study is to establish and verify a advanced control algorithm by analyzing, modelling, stability calculation, controller parameter calculation, simulation for S/G level control system.

• 대한인간공학회:학술대회논문집
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• 대한인간공학회 1993년도 추계학술대회논문집
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• pp.21-30
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• 1993

This paper describes the development of a simulation model of nuclear power plant operators including cognitive aspects by using a network modeling soft ware, Micro-SAINT (System Analysis of Integrated Networks of Tasks) for the analysis of operator performance. Network model description based on Micro-SAINT includes tasks, resources, precedence relations among tasks, flow of information and PSFs (Performance Shaping Factors) on task performance. We have tried to evaluate the performance with several performance measures such as the number of tasks allocated, relative time presure among operators within a shift, for the selected test accident scenarior; small-break LOCA (Loss of Coolant Accident) in a PWR (Pressurized Water Reactor) type nuclear power plant.

• Nuclear Engineering and Technology
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• 제25권1호
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• pp.28-36
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• 1993

최근 국내에서 가압경수로형 원자력발전소를 표준화하기 위한 작업이 이루어지고 있다. 본 논문에서는 설계표준화 작업의 일환으로서 원자력발전소 원자로내부구조물에 대한 내진설계기준을 제시하였다. 영광 3,4호기 최종설계단계에서의 운전기준지진에 대한 원자로용기 플랜지와 스너버의 거동을 입력하중으로 사용하여 지진설계하중을 계산하였고 이로부터 원자로내부구조물의 설계에 허용가능한 원자로용기의 거동을 규정하였다. 해석방법등 해석의 전반적인 개요에 대하여 설명하였고 원자로용기의 거동에 따른 원자로내부구조물 각각의 응답에 대하여 자세히 고찰하였다.

• Nuclear Engineering and Technology
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• 제28권3호
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• pp.247-256
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• 1996

This paper presents a new approach for assessing accident management strategies using containment event trees (CETs) developed during an individual plant examination (IPE) for a reference plant (CE type, 950 MWe PWR). Various accident management strategies to reduce risk have been proposed through IPE. Three strategies for the station blackout sequence are used as an example : 1) reactor cavity flooding only, 2) primary system depressurization only, and 3) doing both. These strategies are assumed to be initiated at about the time of core uncovery. The station blackout (SBO) sequence is selected in this paper since it is identified as one of the most threatening sequences to safety of the reference plant. The effectiveness and adverse effects of each accident management strategy are considered synthetically in the CETs. A best estimate assessment for the developed CETs using data obtained from NUREG-1150, other PRA results, and the MAAP code calculations is performed. The strategies are ranked with respect to minimizing the frequencies of Various containment failure modes. The proposed approach is demonstrated to be very flexible in that it can be applied to any kind of accident management strategy for any sequence.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.21-26
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• 1996

Recently several PWR Nuclear Plant experienced fuel rod fretting wear failures due to Flow Induced Vibration. When such multi-span supported fuel assembly has vibration excitation, it is important to know how fretting wears are progress and when the fuel rods are start to failure. In this study, we estimate the amount of wear depth using Archard theory when the fuel rod starts to relative motion against spacer grid dimples.

• Nuclear Engineering and Technology
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• 제1권1호
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• pp.55-66
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• 1969

1974년도에 가동될 고리 원자력 발전소(Westinghouse 600 MWe PWR)의 핵연료 주기비를 계산했다. 적용한 가정과 가격은 현재 핵 공학계에서 인용하는 것에 기준을 두었고, 한국이라는 국지적 조건을 참작하였다. 계산방식은 가장 적절하다고 생각되는 것을 Normal로 두어, 이 보다 좋거나 나쁜 조건을 고려하였다. 마지막으로 각 Parameter가 주어진 범위 내에서 변하는 것을 Normal과 비교하고 전비용중에서 각각이 차지하는 비율을 검토했는데 그 결과 Uranium 원광비, 성형가공비가 가장 비율이 크고, 그 다음이 이자율, Plutonium Credit, Plant Capacity Factor 등의 순이었다.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1997년도 춘계학술대회논문집; 경주코오롱호텔; 22-23 May 1997
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• pp.488-492
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• 1997

원자력 발전소의 공조 시스템에 설치되는 덕트형 소음기를 설계.제작하여 감음성능 시험, 내진해석, 내진시험, 난연시험 및 기밀시험 등을 수행하여 각각의 요구조건에 적합한 소음기를 설계.제작한 결과 원자력 발전소에 기 설치된 소음기와 동등한 성능을 갖는 것으로 나타났다. 소음기를 설계, 제작, 시험하는 과정에서의 노-하우를 바탕으로 어떠한 요구조건에도 만족하는 제품을 국내에서 설계.생산할 수 있는 기술 확보에 있다.

• 방사성폐기물학회지
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• 제4권1호
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• pp.51-58
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• 2006

고리 원전 사용후핵연료 저장조의 저장용량을 확보하기 위하여 2002년부터 사용후핵연료 운반용기를 이용하여 400다발 이상의 PWR 사용후핵연료 집합체를 원전부지 내에 수송, 저장하였다. 이를 위하여 KN-12 운반용기, 관련장비 및 수송차량으로 구성되는 수송시스템을 구성하였다. KN-12 운반용기는 국내 원자력법 및 IAEA의 수송규정에 따라 설계, 제작되고, 정부로부터 인허가를 획득하였으며, 취급장비 역시 관련규정에 따라 구비하였다. 수송 저장작업은 2 대의 운반용기를 동시에 투입하여 수행하였으며, 모든 작업공정에 대하여 엄격한 품질관리 및 방사선 안전관리를 수행하여 수송 안전성을 확보하고 신뢰도를 제고하였다.