• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2012년도 추계 학술논문 발표대회
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• pp.167-168
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• 2012

Analysis on structural effects which are reduction of friction coefficient and increase of tendon area by HDPE greased and large size strand in post-tensioning system of reactor containment building was carried out. Effective ratio of tendon force increases 67% to 83% by HDPE greased strand and vertical, horizontal internal section forces increased maximum 51%, 41% respectively. Tendon quantity could be reduced 30% by large size and HDPE greased strand that can maintain safety of ultimate internal pressure same as at present.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2012년도 추계 학술논문 발표대회
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• pp.165-166
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• 2012

A thermal stresses by heat of hydration was analyzed according to a change of a pour height in reactor containment building. In case of more than 3.6m pouring height a crack index by heat of hydration analysis resulted in less than 1 because there is not a construction joint of vertical direction and for a self-restraint effect of circumferential section shape. Therefore detailed consideration on a mixture proportion of binder type, quantity in concrete and selection of a form in seasonal air temperature is needed for a control of tensile stress by heat of hydration.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2012년도 추계 학술논문 발표대회
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• pp.169-170
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• 2012

It is analyzed that constructability of post-tensioning system applying HDPE greased strand that is greased and coated by high density polyethylene on a bare strand in reactor containment building. The improvement of corrosion resistance by greasing and HDPE coating on a strand makes transportation, handling and installation of tendon to be easier. Therefore, serial and repetitive process of post-tensioning composed of construction preparation, tendon installation, stressing and anchoring, grease injection could be improved parallel and lumping process of installation and grouting, stressing and anchoring.

• 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.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.37-41
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• 2020

The Nuclear containment building is a main safety-related structure that performs shielding and conservation functions to prevent highly radioactive materials from leakage to the outside environment in the case of various environmental conditions and postulated accidents. The containment building contains a reactor, steam generator, pressurizer, tank, reactor coolant system, auxiliary system and engineering safety system, and is designed so that highly radioactive materials above the limits specified in 10 CFR 100 do not escape to the outside environment in the case of LOCA(Loss of Coolant Accident) for instance. The containment metal liner plate(CLP) is a carbon steel plate with a nominal plate thickness of 6 mm, which functions as a mold for the wall and dome of the containment building when concrete is filled, fulfills airtightness to prevent leakage of seriously radioactive materials. In recent years, backside corrosion was found on the liner plate in some domestic nuclear power plants. The main cause of backside corrosion was unfilled concrete. In this paper, an inspection technique of assessing filling suitability for CLP backside concrete is developed. Results show that the validity of inspection technique for CLP backside concrete using vibration sensor is successfully verified.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4179-4188
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• 2021

This study identifies efficient earthquake intensity measures (IMs) for seismic performances and fragility evaluations of the reactor containment building (RCB) in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). The computational model of RCB is constructed using the beam-truss model (BTM) for nonlinear analyses. A total of 90 ground motion records and 20 different IMs are employed for numerical analyses. A series of nonlinear time-history analyses are performed to monitor maximum floor displacements and accelerations of RCB. Then, probabilistic seismic demand models of RCB are developed for each IM. Statistical parameters including coefficient of determination (R²), dispersion (i.e. standard deviation), practicality, and proficiency are calculated to recognize strongly correlated IMs with the seismic performance of the NPP structure. The numerical results show that the optimal IMs are spectral acceleration, spectral velocity, spectral displacement at the fundamental period, acceleration spectrum intensity, effective peak acceleration, peak ground acceleration, A95, and sustained maximum acceleration. Moreover, weakly related IMs to the seismic performance of RCB are peak ground displacement, root-mean-square of displacement, specific energy density, root-mean-square of velocity, peak ground velocity, Housner intensity, velocity spectrum intensity, and sustained maximum velocity. Finally, a set of fragility curves of RCB are developed for optimal IMs.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2014년도 추계 학술논문 발표대회
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• pp.205-206
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• 2014

It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. The purpose of this study attempts to evaluate the single-stage workability of the system given a change in the height of the setting of RCB exterior wall formwork to be used in nuclear power plant construction. As a result of this study, it is possible height of 3.5m~4m uses formwork when analyzing the construction period and material costs an increase in formwork by concrete lateral pressure, to ensure the workability of the RCB exterior wall formwork. Through this study, I want to provide as basic data for the improvement of workability and RCB shortening the construction period.

• 한국구조물진단유지관리공학회 논문집
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• 제5권2호
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• pp.147-154
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• 2001

The prestressed concrete containment is one of the most important structures in nuclear power plants, which is required to prevent release of radioactive or hazardous effluents to the environment even in the case of a severe accident. Numerical analyses are carried out by using the ABAQUS finite element program to assess the ultimate pressure capacity of the Y prestressed concrete containment with light water reactor at design criteria condition and aging condition considering varied properties of time-dependant materials respectively. From the results, it is verified that the structural capacity of the Y prestressed concrete containment building under the present, aging condition is still robust. In addition, the parameter studies for the reduction of the ultimate pressure capacity of containment building according to the degradation levels of the main structural materials are carried out. The results show that when the degradations of each materials are considered as individual and combined forms, the influence is large in the order of tendon, rebar and concrete degradation, and tendon-rebar, tendon-concrete and rebar-concrete degradation respectively.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.249-260
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• 2012

Main focus of the combined nuclear research activities at Aachen University (RWTH) and the Research Center J$\ddot{u}$lich (J$\ddot{U}$LICH) is the experimental and analytical investigation of containment phenomena and processes. We are deeply convinced that reliable simulations for operation, design basis and beyond-design basis accidents of nuclear power plants need the application of so-called lumped-parameter (LP) based codes as well as computational fluid dynamics (CFD) codes in an indispensable manner. The LP code being used at our institutions is the GRS code COCOSYS and the CFD tool is ANSYS CFX mostly used in German nuclear research. Both codes are applied for safety analyses especially of beyond design accidents. Focal point of the work is containment thermal-hydraulics, but source term relevant investigations for aerosol and iodine behavior are performed as well. To increase the capability of COCOSYS and CFX detailed models for specific features, e.g. recombiner behavior including chimney effect, building condenser, and wall condensation are developed and validated against facilities at different scales. The close connection between analytical and experimental activities is notable and identifying feature of the RWTH/J$\ddot{U}$LICH activities.

• 한국정밀공학회지
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• 제28권2호
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• pp.233-238
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• 2011

In piping design of nuclear power plant facilities, the load stress according to self-weight is important for design values in test run(shutdown and starting). But sometimes it needs more studies, such as seismic analysis of an earthquake of power plant area and fatigue life and stress of thermal expansion and anchor displacement in operating run. In this paper, seismic evaluations were performed to nuclear piping system of Shin-Kori NO. 3&4 being built in Pusan lately. Results of seismic analysis are evaluated on basis of KEPIC MN code. The structural integrity on RCB piping system was proved.