• 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.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.539-552
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• 2013

To establish an aging management plan considering seawater influx and changes in groundwater within nuclear power plant sites, the characteristics of groundwater flow must be understood. This study investigated the characteristics of groundwater flow within the site and analyzed groundwater level recorded by monitoring wells to evaluate groundwater flow characteristics and elements that affected these characteristics for supplying the information to conduct the appropriate aging management for ensuring the safety of the safety-related structures in Shin Kori Unit 1 and 2. The increase in groundwater level during the wet season results from high sea-level conditions and the large amount of precipitation. As a result of the analysis of groundwater distribution and change characteristics, the site could be divided into a rainfall-affected area and a tide-affected area. First, the rainfall-affected area can further be divided into areas that are affected simultaneously by excavation, backfill, and a permanent dewatering system. Secondly, areas that are not affected by excavation, or the dewatering system, or by structure arrangement and excavation. Analysis of the spectrum for wells affected by tides resulted in confirmation of the M2 component (12.421 hr) and S2 component (12.000 hr) of the semidiurnal tides, and the O1 component (25.819 hr) of the diurnal tides. In the cross-correlation results regarding tides and groundwater levels, the lag time occurred diversely within 1-3 hours by the effect of the well location from sea, the distribution of the backfill material with depth, and the concrete structure.

• Journal of Welding and Joining
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• 제31권6호
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• pp.1-7
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• 2013

In nuclear power plants, lined carbon steel pipes or PCCPs (pre-stressed concrete cylinder pipes) have been widely used for sea water transport systems. However, de-bonding of linings and oxidation of PCCP could make problems in aged NPPs (nuclear power plants). Recently at several NPPs in the United States, the PCCPs or lined carbon steel pipes of the sea water or raw water system have been replaced with HDPE (high density polyethylene) pipes, which have outstanding resistance to oxidation and seismic loading. ASME B&PV Code committee developed Code Case N-755, which describes rules for the construction of buried Safety Class 3 polyethylene pressure piping systems. Although US NRC permitted HDPE materials for Class 3 buried piping, their permission was limited to only 10-year operation because of several concerns including the quality of fusion zone of HDPE. In this study, various requirements for fusion qualification test of HDPE and some regulatory issues raised during HDPE application review in foreign NPPs are introduced.

• 한국지진공학회논문집
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• 제17권2호
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• pp.53-59
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• 2013

Several researches have been studied to enhance the seismic performance of nuclear power plants (NPPs) by application of seismic isolation. If a seismic base isolation system is applied to NPPs, seismic performance of nuclear power plants should be reevaluated considering the soil-structure interaction effect. The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP structures and equipment. In this study, the seismic performance of an isolated NPP is evaluated by seismic fragility curves considering the soil-structure interaction effect. The designed seismic isolation is introduced to a containment building of Shin-Kori NPP which is KSNP (Korean Standard Nuclear Power Plant), to improve its seismic performance. The seismic analysis is performed considering the soil-structure interaction effect by using the linearized model of seismic isolation with SASSI (System for Analysis of Soil-Structure Interaction) program. Finally, the seismic fragility is evaluated based on soil-isolation-structure interaction analysis results.

• Journal of Radiation Protection and Research
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• 제9권2호
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• pp.118-128
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• 1984

1978년(年) 4월(月) 29일(日) 상업운전을 개시한 원자력(原子力) 1 호기(號機)의 방사선관리실태放射線管理實態)를 방사선피폭(放射線被爆), 방사성폐기물(放射性廢棄物) 및 환경관리(環境管理)로 구분고찰(區分考察) 하였다. 그 결과(結果) 방사선피폭(放射線被爆)과 방사성폐기물(放射性廢棄物)은 발전소(發電所) 운전년수(運轉年數)에 따라 점차(漸次) 증가현상(增加現象)을 나타내더니, 1983 및 1982 년도(年度)를 정점(頂點)으로 감소(減少) 또는 안정추세(安定趨勢)를 나타냈다. 그리고 발전소 가동후(稼動後)의 주변환경방사능(周邊環境放射能)도 발전소 가동전(稼動煎)보다 높지 않은것으로 나타났다.

• 한국건축시공학회지
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• 제15권4호
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• pp.425-431
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• 2015

1978년에 고리 1호기의 건설이 완공된 이래로 원자력 발전 플랜트의 건설 프로젝트는 국내 외로 점차 확대되고 있다. 그러나 일부 원자력 발전 플랜트의 건설 현장에서는 리스크 관리 능력의 부족으로 인하여 공기 지연과 공사비 손실의 문제점들을 가지고 있다. 특히, 원자력 발전 플랜트 내 원자로격납건물의 시공은 타 시공 단계에 비해 긴 공정기간으로 인하여 전문기술과 대규모 자원이 요구됨에 따라 많은 리스크 요인들이 산재될 수 있다. 따라서 원자로격납 건물의 시공에서 예상되는 리스크 요인들을 분석하여 전체 프로젝트의 안정적인 수행 방향을 제시하는 연구가 필요하다. 그러므로 본 연구는 원자로격납건물 시공의 리스크 요인들을 평가하고자 한다. 이를 위하여 본 연구는 36명의 소수 전문가 집단을 대상으로 하는 설문조사방법을 활용하였다. 24개의 리스크 요인들은 공정, 원가, 안전, 품질을 기준으로 분류되었으며, 이에 대한 평가 결과는 계층 분석방법을 활용하여 분석하였다. 이를 바탕으로 각 기준별로 분류된 리스크 요인들은 중요도와 우선순위를 산정하고 원자력 발전 플랜트의 시공 리스크 요인을 분석하는데 계층분석 방법의 적용성을 확인하였다. 본 연구의 결과는 원자로격납건물의 시공 단계에서 리스크 관리를 위한 기초 자료로 활용될 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제42권2호
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• pp.219-230
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• 2010

In nuclear power plants (NPPs), periodic sensor calibrations are required to assure that sensors are operating correctly. By checking the sensor's operating status at every fuel outage, faulty sensors may remain undetected for periods of up to 24 months. Moreover, typically, only a few faulty sensors are found to be calibrated. For the safe operation of NPP and the reduction of unnecessary calibration, on-line instrument calibration monitoring is needed. In this study, principal component-based auto-associative support vector regression (PCSVR) using response surface methodology (RSM) is proposed for the sensor signal validation of NPPs. This paper describes the design of a PCSVR-based sensor validation system for a power generation system. RSM is employed to determine the optimal values of SVR hyperparameters and is compared to the genetic algorithm (GA). The proposed PCSVR model is confirmed with the actual plant data of Kori Nuclear Power Plant Unit 3 and is compared with the Auto-Associative support vector regression (AASVR) and the auto-associative neural network (AANN) model. The auto-sensitivity of AASVR is improved by around six times by using a PCA, resulting in good detection of sensor drift. Compared to AANN, accuracy and cross-sensitivity are better while the auto-sensitivity is almost the same. Meanwhile, the proposed RSM for the optimization of the PCSVR algorithm performs even better in terms of accuracy, auto-sensitivity, and averaged maximum error, except in averaged RMS error, and this method is much more time efficient compared to the conventional GA method.

• 방사성폐기물학회지
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• 제17권3호
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• pp.299-311
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• 2019

최근 한국에서 원전해체는 중요한 이슈이다. 원전의 운영 시와 비교해볼 때, 원전 해체 시에는 방사성물질의 방출이 크지 않을 것으로 예상되지만, 주민은 항상 방사선피폭으로부터 보호되어야 한다. 이에 대한 효과적인 관리를 위해, 연간 방출관리치와 방출한도치를 원자력안전위원회 고시 및 일반인 선량한도 기준으로부터 유도하였다. 기체상 유출물에 의한 대기 확산 및 침적 인자는 신고리 발전소 기상탑에서 2008년부터 2010년까지 3년간 수집 된 기상자료를 토대로 XOQDOQ 컴퓨터 코드를 이용해서 평가하였다. 선량평가는 ENDOS-G 컴퓨터 코드를 사용하였다. 이 컴퓨터 코드를 이용하여 기체상 유출물의 연간 방출관리치 및 방출한도치를 평가한 결과, 핵종별로 차이가 있었는데, 이는 연령에 따른 방사선민감도의 차이에 기인한다고 할 수 있다. 본 평가 방법 및 결과는 향후 원전 해체 시 방사성유출물 관리에 중요한 정보를 제공할 수 있을 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.234-243
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• 2021

In the research or project planning for the decommissioning of a nuclear power plant, one of several preparations will be the establishment of a list of potential radionuclides to be considered at the time of characterization or final status surveys. Reliable data for selection of potential radionuclides during the transition period to prepare for decommissioning will depend heavily on historical data at the site or, where possible, sampling analysis. However, during the transition period, direct sampling can be challenging, depending on the circumstances of the site or national regulation. A methodology of selecting potential radionuclides for nuclear facility sites which largely consists of three major processes: production of initial list of radionuclides, selection of the insignificant radionuclide that will be eliminated, and consideration of site characterization or sampling. For developing a preliminary list of potential radionuclides for Kori Unit 1 decommissioning, the list of initial radionuclides was made referring to the technical documents applied at decommissioned NPPs in the U.S and additional reference materials applied until the operation of NPPs in Korea. For the screening of insignificant radionuclides, we applied criterion of less than 0.1% of the amount of radioactivity inventory and confirmed the dose fraction using the RESRAD code. The final suit of radionuclides was established, which should be supplemented by reflecting site characterization and sampling process in the future. Thus, the methodology and results for the selection of potential radionuclides suggested in this paper can give an insight as a future reference to deriving DCGLs in relation to site remediation of decommissioning nuclear plants.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.196-205
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• 2016

Background: The emergency planning zone (EPZ) of the city of Busan is divided into the precautionary actions zone (PAZ) and the urgent protective action planning zone; which have a 5-km radius and a 20-km to 21-km radius from the nuclear power plant site, respectively. In this study, we assumed that a severe accident occurred at Shin-Kori nuclear unit 3 and evaluated the dispersion speed of radiological material at each distance at various wind speeds, and estimated the effective dose equivalent and the evacuation time of PAZ residents with the goal of supporting off-site emergency action planning for the nuclear site. Materials and Methods: The total effective dose equivalent, which shows the effect of released radioactive materials on the residents, was evaluated using the RASCAL 4.2 program. In addition, a survey of 1,036 residents was performed using a standardized questionnaire, and the resident evacuation time according to road and distance was analyzed using the VISSIM 6.0 program. Results and Discussion: According to the results obtained using the VISSIM and RASCAL programs, it would take approximately 80 to 252.2 minutes for permanent residents to move out of the PAZ boundary, 40 to 197.2 minutes for students, 60 to 232.2 minutes for the infirm, such as elderly people and those in a nursing home or hospital, and 30 to 182.2 minutes for those temporarily within the area. Consequently, in the event of any delay in the evacuation, it is estimated that the residents would be exposed to up to $10mSv{\cdot}h^{-1}$ of radiation at the Exclusion Area Boundaries (EAB) boundary and $4-6mSv{\cdot}h^{-1}$ at the PAZ boundary. Conclusion: It was shown that the evacuation time for the residents is adequate in light of the time lapse from the initial moment of a severe accident to the radiation release. However, in order to minimize the evacuation time, it is necessary to maintain a system of close collaboration to avoid traffic congestion and spontaneous evacuation attempts.