• Journal of Radiation Protection and Research
• /
• v.31 no.2
• /
• pp.53-63
• /
• 2006

The increased risk of cancer with exposure to low-dose radiation was estimated through longitudinal study for radiation workers at the nuclear power plants in Korea. The radiation dose data were collected from the Radiation Safety Management System(RSMS) of the Korea Hydro & Nuclear Power Co., Ltd(KHNP). The cancer risks with radiation exposure were evaluated in terms of relative mortality ratios(RMR) and relative incidence ratios(RIR) to the unexposed employees at the nuclear power plants, and of the standardized mortality ratios(SMR) and standardized incidence ratios(SIR). There were no significant increases of canters of all sites in the exposed group either in RIR[1.08, 95% confidence interval(CI) 0.74-1.58] or in RMR[1.21, CI 0.70-2.08]. Neither SIR[0.81, CI 0.28-0.96] nor SMR[0.86, CI 0.66-1.10] significantly deviated from 1.0 for cancers of all sites. The trend analysis did not identify evident dose-response relationship due to insufficient numbers of the cases. Consequently, it is concluded that increases in cancer risks in the radiation worker group exposed to low doses at the nuclear power plants in Korea are not identified at this time.

• Nuclear Engineering and Technology
• /
• v.52 no.4
• /
• pp.764-775
• /
• 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.

• Proceedings of the KIEE Conference
• /
• 1999.07c
• /
• pp.1221-1223
• /
• 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.

• Korean Security Journal
• /
• no.61
• /
• pp.9-38
• /
• 2019

As the scales & density of the Korean national infrastructures have been increased, they will be identified as rich and attractive potential targets for intensified North Korea's attack in the rear region and terrorism attack. In addition, due to changes in security environment such as drone threats and lack of security forces under the 52-hour workweek law, I think that it is the proper time point to reevaluate the effectiveness and appropriateness of the current physical protection system and its shift to a new system. In this study, the direction and improvement of the perimeter physical protection systems of the national infrastructures are to be studied from the viewpoints of its concepts of operations and design methodology, focusing on the nuclear power plant. The reason why we focus on nuclear power plants is because they cause wide-range and long-term damages caused by radioactive materials disperal and pollution, along with short-term damage caused by the interruption of electricity generation in the event of damage to nuclear power plants. With the aim of extracting improvement directions, as we will comprehensively review domestic research trends and domestic·overseas related laws, and consider Korea's specificity, we try to reframe the concept of operation - systematization, mobilization and flexibility -, and establish criteria on system change. In order to improve the technical performance of the new perimeter physical protection system, we study on high-fidelity·multi-methodology based integrated design methodology, breaking from individual silo-type design methods, and I suggest improvement of government procurement, its expansion to export business and other national infrastructure.

• Proceedings of the Korean Society of Machine Tool Engineers Conference
• /
• 1996.10a
• /
• pp.107-112
• /
• 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.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 1996.11a
• /
• pp.469-473
• /
• 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.

• The Transactions of the Korean Institute of Electrical Engineers D
• /
• v.55 no.7
• /
• pp.303-312
• /
• 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.

• Journal of the Korean Society of Manufacturing Technology Engineers
• /
• v.6 no.4
• /
• pp.9-18
• /
• 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.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.153-158
• /
• 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).

• Journal of the Korean Society of Safety
• /
• v.35 no.1
• /
• pp.107-115
• /
• 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.