• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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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.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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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.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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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.

• Journal of the Korea Institute of Building Construction
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• v.15 no.4
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• pp.425-431
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• 2015

Since the construction of Kori 1 was completed in 1978, the construction projects for nuclear power plant are increasingly expanded into domestic and foreign sites. However, some of construction sites of nuclear power plant have the problems of process delay and cost loss due to lack of ability of risk management. The construction of reactor containment building in nuclear power plant is especially dotted with many risk factors because it needs professional skills and large-scale resources due to long duration compared with different construction phase. Therefore, it needs the study that analyzes risk factors expected in construction of reactor containment building and suggests way of stable performance of projects. So, this study assesses risk factors of construction of reactor containment building. For the objectives, this study uses survey for group of minority specialists of 36 experts. The risks of 24 factors is classified by criterions of process, cost, safety, and quality and the results of assessment is analyzed by analytic hierarchy process. As the results, the importance and priority of risk factors classified by each criterion were calculated and the applicability of analytic hierarchy process was identified to analyze risk factors of nuclear power plant construction. These will be baseline data for risk management in construction phase of reactor containment building.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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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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• v.53 no.12
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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.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2014.11a
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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.

• Journal of the Korea institute for structural maintenance and inspection
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• v.5 no.2
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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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• v.44 no.3
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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.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3043-3066
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• 2024

In various small modular reactor (SMR) designs currently under development, the conventional concrete containment building has been replaced by a metal containment vessel (MCV). In these systems, the gap between the MCV and the reactor pressure vessel is filled with gas or vacuumed weakly, effectively suppressing conduction and convection heat transfer. However, thermal radiation remains the major mode of heat transfer during normal operation. The objective of this study was to investigate the heat-transfer mechanisms in integral pressurized water reactor (IPWR)-type SMRs under various gas-filled conditions using computational fluid dynamics. The use of thermal radiation shielding (TRS) with a much lower emissivity material than the MCV surface was also evaluated. The results showed that thermal radiation was always the dominant contributor to heat loss (48-97%), while the conjugated effects of the gas candidates on natural convection and thermal radiation varied depending on their thermal and radiative properties, including absorption coefficient. The TRS showed an excellent insulation performance, with a reduction in the total heat loss of 56-70% under the relatively low temperatures of the IPWR system, except for carbon dioxide (13%). Consequently, TRS can be utilized to enhance the thermal efficiency of SMR designs by suppressing the heat loss through the MCV.