• Journal of Radiation Protection and Research
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• 제31권2호
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• pp.53-63
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• 2006

원전 종사자를 대상으로 경시적인 코호트 연구를 통하여 저준위 방사선 노출과 암 발생 위해도를 규명하고자 하였다. 방사선 노출에 관한 정보는 한국수력원자력(주)의 방사선 관리 DB에서 수집하였고, 암 발생에 관한 정보는 한국인 중앙 암 등록 자료를, 암 사망에 관한 정보는 통계청 사망원인 자료를 이용하여 수집하였다. 방사선 노출과 암 발생 위해도는 표준화 암 사망비(SMR)와 표준화 암 발생 비(SIR)로 평가하였다. 노출 군에서 암 발생에 대한 상대위험도는 1.08로 평가되었으나 전체 암에 대한 SIR은 0.81로 유의성이 관찰되지 않았다. 암 사망에 대한 상대위험도는 1.21 이었으나 전체 암에 대한 SMR도 0.86으로 역시 유의성은 관찰되지 않았다. 암 유형별 양상은 우리나라 일반인과 유사한 결과를 보였으며, 방사선량 증가에 따른 양-반응 관계 또한 확인되지 않았다.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.764-775
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• 2020

Periodic safety reviews (PSRs) are conducted on operating nuclear power plants (NPPs) and have been mandated also for research reactors in Korea, in response to the Fukushima accident. One safety review tool, the probabilistic safety assessment (PSA), aims to identify weaknesses in the design and operation of the research reactor, and to evaluate and compare possible safety improvements. However, the PSA for research reactors is difficult due to scarce data availability. An important element in the analysis of research reactors is the reactor protection system (RPS), with its functionality and importance. In this view, we consider that of the AGN-201K, a zero-power reactor without forced decay heat removal systems, to demonstrate a risk-informed safety improvement study. By incorporating risk- and safety-significance importance measures, and sensitivity and uncertainty analyses, the proposed method identifies critical components in the RPS reliability model, systematically proposes potential safety improvements and ranks them to assist in the decision-making process.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1999년도 하계학술대회 논문집 C
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• pp.1221-1223
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• 1999

The safety loads in a nuclear power plant perform a critical function to plant safety. The design of the electrical auxiliary system should ensure the availability and adequacy of the power supply, and therefore, the frequency and voltage relaying schemes should be installed on the system to monitor and protect against the degraded system condition. If unforeseen contingencies degrade the switchyard frequency and voltage to below the minimum values, the safety related bus should properly be transferred to alternate power source. This paper presents guidelines associated with the protection of nuclear power plants during frequency/voltage decay and the steady-state and dynamic analysis of auxiliary power system that should be performed to support the degraded voltage relay(second level undervoltage relay) setting.

• 시큐리티연구
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• 제61호
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• pp.9-38
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• 2019

한국의 국가기반시설은 시설규모가 증가하고 밀집되어 강화된 북한의 국지도발, 테러공격을 위한 풍부하고 매력적인 잠재적 표적으로 식별될 것이다. 또한 드론위협, 주 52시간 근무제도에 따른 경비병력 부족 등의 보안환경 변화에 따라 현 물리적 방호체계에 대한 유효성과 적절성을 재평가하고 전환을 고려할 시점으로 사료된다. 본 연구에서는 국가기반시설 중 원자력발전소의 외곽 물리적 방호체계에 집중하여 국가 기반시설 외곽 물리적 방호체계의 전환 방향과 개선방안을 운영개념 및 설계 방법론 측면에서 연구하였다. 원자력발전소에 집중하는 이유는 원자력발전소는 피해 시 전기발전 중단의 단기적인 피해와 함께, 방사능 물질 유출과 오염에 따르는 광범위하고 장기적인 피해가 발생하므로 가장 높은 보안수준을 필요로 하기 때문이다. 개선방향 도출 목표로 국내 연구동향과 국내·해외 관련법을 종합 검토하고 한국의 특수성을 고려하여, 과학화, 기동화, 유연성으로 운영개념을 재설정하고 체계전환의 기준을 수립하였다. 새로운 외곽 물리적 방호체계의 기술적 성능개선을 위하여 개별설계에서 탈피, 고신뢰성·다방법론 기반의 통합설계 방법론 적용방안을 연구하고 구매제도 개선 및 해외 수출, 타(他)국가기반시설로의 확대적용을 제언한다.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 1996년도 추계학술대회 논문
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• pp.107-112
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• 1996

In this paper it is proposed a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. In nuclear power plants workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from the radiation exposure. Digital signal processors are used in implementing real time recognition and examination of steam generator tubes in the preposed vision system, Performance of proposed digital vision system is illustrated by experiment for similar steam generator model.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 1996년도 추계학술대회 논문집
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• pp.469-473
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• 1996

In this paper, it is proposed a new approach to the development of the automatic vision system to e famine and repair the steam generator tubes at remote distance. In nuclear power plants, workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from the radiation exposure. Digital signal processors are used it, implementing real time recognition and examination of steam generator tubes in the proposed vision system. Performance of proposed digital vision system is illustrated by experiment for similar steam generator model.

• 대한전기학회논문지:시스템및제어부문D
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• 제55권7호
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• pp.303-312
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• 2006

This paper proposes the architecture and protocol of a data communication network for the safety system in nuclear power plants. First, we establish four design criteria with respect to determinability, reliability, separation and isolation, and verification/validation. Next we construct the architecture of the safety network for the following systems: PPS (Plant Protection System), ESF-CCS (Engineered Safety Features-Component Control System) and CPCS (Core Protection Calculator System). The safety network consists of 12 sub-networks and takes the form of a hierarchical star. Among 163 communication nodes are about 1600 origin-destination (OD) pairs created on their traffic demands. The OD pairs are allowed to exchange data only during the pre-assigned time slots. Finally, the communication protocol is designed in consideration of design factors for the safety network. The design factors include a network topology of star, fiber-optic transmission media, synchronous data transfer mode, point-to-point link configuration, and a periodic transmission schedule etc. The resulting protocol is the modification of IEEE 802.15.4 (LR-WPAN) MAC combined with IEEE 802.3 (Fast Ethernet) PHY. The MAC layer of IEEE 802.15.4 is simplified by eliminating some unnecessary (unctions. Most importantly, the optional TDMA-like scheme called the guaranteed time slot (GTS) is changed to be mandatory to guarantee the periodic data transfer. The proposed protocol is formally specified using the SDL. By performing simulations and validations using Telelogic Tau SDL Suite, we find that the proposed safety protocol fits well with the characteristics and the requirements of the safety system in nuclear power plants.

• 한국생산제조학회지
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• 제6권4호
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• pp.9-18
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• 1997

It is proposed a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. In nuclear power plants, workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from to radiation exposure. Digital signal processors are used in implementing real time recognition and examination of steam generator tubes in the proposed vision system. Performance of proposed digital vision system is illustrated by simulation and experiment for similar steam generator model.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.153-158
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• 1995

All of the nuclear power plants in Korea we operating with analog instrumentation and control (I&C) equipment which are increasingly faced with frequent troubles, obsolescence and high maintenance expenses. Electrical and computer technology has improved rapidly in recent years and has been applied to other industries. So it is strongly recommended we adopt modern digital and computer technology to improve plant safety and availability. The advanced I&C system, namely, Integrated Intelligent Instrumentation and Control System (I$^3$CS) will be developed for beyond the next generation nuclear power plant. I$^3$CS consists of three major parts, the advanced compact workstation, distributed digital control and protection system including Automatic Start-up/shutdown Intelligent Control System (ASICS) and the computer-based alarm processing and operator support system, namely, Diagnosis, Response, and operator Aid Management System (DREAMS).

• 한국안전학회지
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• 제35권1호
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• pp.107-115
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• 2020

This paper introduces the first vital area identification (VAI) process for the physical protection of nuclear power plants (NPPs) during low power and shutdown (LPSD) operation. This LPSD VAI is based on the 3rd generation VAI method which very efficiently utilizes probabilistic safety assessment (PSA) event trees (ETs). This LPSD VAI process was implemented to the virtual NPP during LPSD operation in this study. Korea Atomic Energy Research Institute (KAERI) had developed the 2nd generation full power VAI method that utilizes whole internal and external (fire and flooding) PSA results of NPPs during full power operation. In order to minimize the huge burden of the 2nd generation full power VAI method, the 3rd generation full power VAI method was developed, which utilizes ETs and minimal PSA fault trees instead of using the whole PSA fault tree. In the 3rd generation full power VAI method, (1) PSA ETs are analyzed, (2) minimal mitigation systems for avoiding core damage are selected from ETs by calculating system-level target sets and prevention sets, (3) relatively small sabotage fault tree that has the systems in the shortest system-level prevention set is composed, (4) room-level target sets and prevention sets are calculated from this small sabotage fault tree, and (5) the rooms in the shortest prevention set are defined as vital areas that should be protected. Currently, the 3rd generation full power VAI method is being employed for the VAI of Korean NPPs. This study is the first development and application of the 3rd generation VAI method to the LPSD VAI of NPP. For the LPSD VAI, (1) many LPSD ETs are classified into a few representative LPSD ETs based on the functional similarity of accident scenarios, (2) a few representative LPSD ETs are simplified with some VAI rules, and then (3) the 3rd generation VAI is performed as mentioned in the previous paragraph. It is well known that the shortest room-level prevention sets that are calculated by the 2nd and 3rd generation VAI methods are identical.