• 조명전기설비학회논문지
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• 제25권4호
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• pp.71-78
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• 2011

This paper contains the selection of operation parameters for protection and control of steam turbine in a Korea Standard Type Nuclear Power Plant. The safety of nuclear reactor must be ensured which generates nuclear energy and produces steam. Also, the safety of turbine, which consume the nuclear energy as a core machine, must be ensured. For the purpose of this, we describe how the operating parameters were selected, reviewed, implemented into the operator console and finally put into actual operation of the system.

• 국제학술발표논문집
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• The 7th International Conference on Construction Engineering and Project Management Summit Forum on Sustainable Construction and Management
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• pp.8-13
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• 2017

Nuclear power plant (NPP) constructions are large scale projects that are executed for several years, and schedule control utilizing various schedules is a critically important factor. Recently Korea independently developed the Advanced Power Reactor (APR) 1400 and is building nuclear facilities applying this new reactor type. The construction of Shin-Kori NPP (SKN) Unit 3, which adopted the APR1400, was completed and commercial operation has begun, while, SKN 4, Shin-Hanul NPP (SHN) Units 1&2, and SKN 5&6 are currently under construction. Prior to the development of the APR1400, Korea built 24 reactors and accumulated the schedule data of various reactor types which provided the foundation for schedule reduction to be possible. However, as there is no schedule development and review system established based on the standard schedule data (standard activities, durations, etc.) by reactor type, the process for developing the schedule for new builds is low in efficiency consuming much time and manpower. Also all construction data has been accumulated based on schedule activities. But because the connectivity of activities between projects is low, it is difficult to utilize such accumulated data (causes for schedule delay, causes for design changes, etc.) in new build projects. Due to such reasons, issues continue to arise in the process of developing standard schedule activities and a standard schedule for nuclear power plant construction. In order to develop a standard schedule for NPP construction, i) the development of an NPP standard schedule activity list, ii) development of the connection logic of NPP standard schedule activities, iii) development of NPP standard schedule activity resources and duration, and iv) integration of schedule data need to be performed. In this paper, an analysis was made on the coherence of schedule activity descriptions of existing NPPs by applying the probabilistic methodology on activities with low connectivity due to the utilization of the numbering system of four APR1400 reactors (SHN 1&2 and SKN 3&4).This study also describes the method for developing a standard schedule activity list and connectivity measures by extracting same and/or similar schedule activities.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.244-250
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• 1996

Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth.

• Journal of Radiation Protection and Research
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• 제19권1호
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• pp.23-35
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• 1994

CANDU형 원자력 발전소에서의 유도방출한도를 계산하기 위한 전산 코드 DRL이 개발되었다. DRL 코드에서의 유도방출한도는 CANDU형 원자력 발전소가 정상 가동될 때의 기체 및 액체 방출물에 포함된 방사성 핵종의 방출 허용 기준을 설정하기 위한 것이다. 본 전산 코드는 CSA Standard N288.1-M87에서 권고하고 있는 방법 및 다수 매개 변수를 이용하였고, 월성 원자력 발전소를 대상으로 유도방출한도를 결정하는데 이용되었다.

• Journal of Radiation Protection and Research
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• 제34권2호
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• pp.77-81
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• 2009

An Area Radiation Monitoring System (ARMS) ionization chamber, which had an 11.8 L active volume, was fabricated and performance-tested at KAERI. Low leakage currents, linearities at low and high dose rates were achieved from performance tests. The correlation coefficients between the ionization currents and the dose rates are 1 at high dose rate and 0.99 at low dose rate. In this study, an integration-type ARMS ionization chamber was tested over a year for an evaluation of its long-term stability at a radioisotope (RI) repository of the Young-gwang nuclear power plant. The standard deviation of dose rate of 1 day data and over a 100-days mean value were 6.2 $\mu$R/h and 2.9 $\mu$R/h, respectively. The fabricated ARMS ionization chamber showed stable performance from the results of the long-term tests. Design and performance characteristics of the fabricated ionization chamber for the ARMS from performance-tests are also addressed.

• 한국안전학회지
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• 제27권5호
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• pp.224-228
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• 2012

According to the results of Probabilistic Safety Assessment(PSA) for a Nuclear Power Plant(NPP), an Emergency Power Supply(EPS) system has been considered as one of the most important safety system. Especially, the interests in the reliability of the EPS system have been increased after the severe accidents of Fukushima Daiichi. Firstly, we performed the risk assessment and the importance analysis of the EPS system based on the PSA models of the reference plant, which is the Korean standard NPP type. Considering a portable Diesel Generator(DG) system as the reliability reinforcement of the EPS system, we modified the PSA models and performed the risk impact assessment and the importance analysis. Although the reliability of the potable DG could be about 20% of the reliability of the alternative AC DG, we identified that Core Damage Frequency(CDF) was decreased by at least 4.6%. In addition, the risk impacts due to the unavailability of the EPS system on CDF were decreased.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.497-511
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• 2004

A PSV (pressurizer safety valve) popping test carried out in the early phases of a refueling outage may trigger a test-induced LOCA(loss of coolant accident) if a PSV fails to fully close and is stuck in a partially open position. According to a KSNP (Korea standard nuclear power plant) low power and shutdown PSA (probabilistic safety assessment), the failure of a high pressure safety injection (HPSI) accompanied by the failure of a PSV to fully close was identified as a dominant accident sequence with a significant impact on low power and shutdown risks (LPSR). In this study, we aim to investigate and verify a new means for mitigating this type of accident using a thermal-hydraulic analysis. In particular, we explore the applicability of an aggressive cool-down combined with operator actions. The results of the various sensitivity studies performed there will help reduce LPSR and improve Refueling outage safety.

• 한국안전학회지
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• 제21권5호
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• pp.77-83
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• 2006

The purpose of this study is to assess the organizational personality types, employee-organization fits and the job satisfaction/involvement in a Korea standard nuclear power plant(NPP), which is a representative safety work place. First we chose 427 procedures that are related to safety out of 777 officially managed procedures referenced by 13.5 of FSAR(final safety analysis report). Next, we finally chose 70 procedures of 8 divisions for 44 employees regarding the duties for NPPs' division, experiences of operations, an operational know-how, and the indication of operational weakness. This study used OPTI(organizational personality type indicators) and the combination of 4 preference types for determining the organizational personality to produce personality types of organizations for NPPs' division. To assess the job satisfaction and involvement, we used a questionnaire and an interview, for 300 employees(83.5%) of the Korea standard NPP.

• 한국시뮬레이션학회논문지
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• 제19권1호
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• pp.41-51
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• 2010

500MW급 초임계압 표준 석탄 화력 발전소에의 실증 적용을 목표로 국산 분산제어시스템의 개발 연구가 진행되고 있다. 시뮬레이터는 원자력발전소의 디지털 제어시스템 업그레이드나 아날로그 제어 시스템의 디지털 제어시스템으로의 개체시 제어 시스템의 검증용으로 활용되고 있으며, 국내에서도 중용량 석탄 화력 보일러 제어 시스템 검증에 활용된 바 있다. 본 논문에서 는 500MW급 표준 석탄 화력 발전소에의 적용을 목적으로 개발 중인 제어 시스템을 검증하기 위해 제어 검증용 시뮬레이터를 개발하였다. 제어 검증을 위한 제어 모델을 개발하는데 있어서 현장 제어 시스템 데이터와 시뮬레이션 개발 환경에서 현장 데이터를 활용가능하게 하는 변환 프로그램, 제어 시스템 제작사 매뉴얼에 기반하여 제어 모델을 개발하였다. 개발한 시뮬레이터는 열평형상태시험, 부하변동 시험, 고장모사시험 등을 통하여 효용성을 확인하였으며, 개발 중인 제어 시스템을 검증하고, 기존 제어 시스템의 분석 및 개선에 유용하게 활용할 수 있을 것으로 기대한다.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3353-3360
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• 2022

Using data from the design earthquake of the Korean standard nuclear power plant, seismic analyses of a fuel assembly are conducted in this study. The modal characteristics are used to develop an input deck for the seismic analysis. With a time history analysis, the responses of the fuel assembly in the event of an earthquake are obtained. In particular, the displacement, velocity, and acceleration responses at the center location of the fuel assembly are obtained in the time domain, with these outcomes then used for a detailed structural analysis of the fuel rods in the ensuing analyses. The response spectra are also generated to determine the response characteristics in the frequency domain. The structural integrity of the fuel assembly can be ensured through this type of time history analysis considering the input excitations of various earthquakes considered in the design.