• KSCE Journal of Civil and Environmental Engineering Research
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• v.26 no.5A
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• pp.901-908
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• 2006

The structural integrity of containment building in Nuclear Power Plants has to be verified by the ISI(In Service Inspection) because there are some variations on the structural behavior of it due to the change of the physical properties of concrete and tendon with the lapse of time. In this study, the program 'SAPONC-CANDU' which can monitor and analyze the structural behavior of the containment building with grouted tendon (CANDU-type, 'Wolsong unit-2, 3, and 4' in Korea) was developed. This program is based on the algorithm which can calculate the prediction values of the quantities of strain variation for the vibrating-wire strain gauges embedded into the concrete of the containment building under temperature and time dependent factors which are creep, shrinkage, and prestressing force. The readings of the strain gauges are used as input data for the operation of the program. And it finally provides graphically a prediction value, line and band of the quantity of strain variation for the respective strain gauges, therefore, it is thought that the site engineers are able to assess the structural integrity of the domestic containment building with grouted tendon with easy using this program.

• Earthquakes and Structures
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• v.15 no.3
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• pp.295-306
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• 2018

Reinforced concrete containment (RCC) building has long been considered as the last barrier for keeping the radiation from leaking into the environment. It is important to quantify the performance of these structures and facilities considering extreme conditions. However, the preceding research on evaluating nuclear power plant (NPP) structures, particularly considering mainshock-aftershock seismic sequences, is deficient. Therefore, this manuscript serves to investigate the seismic fragility of a typical RCC building subjected to mainshock-aftershock seismic sequences. The implementation of the fragility assessment has been performed based on the incremental dynamic analysis (IDA) method. A lumped mass RCC model considering the tri-linear skeleton curve and the maximum point-oriented hysteretic rule is employed for IDA analyses. The results indicate that the seismic capacity of the RCC building would be overestimated without taking into account the mainshock-aftershock effects. It is also found that the seismic capacity of the RCC building decreases with the increase of the relative intensity of aftershock ground motions to mainshock ground motions. In addition, the effects of artificial mainshock-aftershock ground motions generated from the repeated and randomized approaches and the polarity of the aftershock with respect to the mainshock on the evaluation of the RCC are also researched, respectively.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.3
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• pp.231-235
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• 2016

There are two methods for conducting the containment local leakage rate test (LLRT) in nuclear power plants: the make-up flow rate method and the pressure decay method. The make-up flow rate method is applied first in most power plants. In this method, the leakage rate is measured by checking the flow rate of the make-up flow. However, when it is difficult to maintain the test pressure because of excessive leakage, the pressure decay method can be used as a complementary method, as the leakage rates at pressures lower than normal can be measured using this method. We studied the method of measuring over leakage using the pressure decay method for conducting the LLRT for the containment building at a nuclear power plant. We performed experiments under conditions similar to those during an LLRT conducted on-site. We measured the characteristics of the leakage rate under varies pressure decay conditions, and calculated the compensation ratio based on these data.

• Journal of the Earthquake Engineering Society of Korea
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• v.25 no.2
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• pp.83-91
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• 2021

Seismic designs for Korean nuclear power plants (NPPs) under earthquakes' design basis are noticed due to the recent earthquake events in Korea and Japan. Japan has developed the technologies and experiences of the NPPs through theoretical research and experimental verification with extensively accumulated measurement data. This paper describes the main features of the design-time history complying with the Japanese seismic design standard. Proper seed motions in the earthquake catalog are used to generate one set of design time histories. A magnitude and epicentral distance specify the amplitude envelope function configuring the shape of the earthquake. Cumulative velocity response spectral values of the design time histories are compared and checked to the target response spectra. Spectral accelerations of the time histories and the multiple-damping target response spectra are also checked to exceed. The generated design time histories are input to the reactor building seismic analyses with fixed-base boundary conditions to calculate the seismic responses. Another set of design time histories is generated to comply with Korean seismic design procedures for NPPs and used for seismic input motions to the same reactor containment building seismic analyses. The responses at the dome apex of the building are compared and analyzed. The generated design time histories will be also applied to subsequent seismic analyses of other Korean standard NPP structures.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.23 no.10
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• pp.1274-1284
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• 1999

The Regulatory Guide 1.82 recommends an analysis of hydraulic performance for sump of ECCS (Emergency Core Cooing System) when LOCA(Loss of Coolant Accident) occurs in a nuclear power plant. The present study deals with 3-dimensional, unsteady, turbulent and two-phase flow simulation to examine the behavior of mixture of reactor coolant and debris near the floor of containment building in conjunction with appropriate assumptions. The dispersed solid model has been adjusted to the interfacial momentum transfer between reactor coolant and debris. According to the results, the counterclockwiserecirculation zone had been formed in the region between sump and connection aisle about 376 second after LOCA occurs. The debris thickness accumulated on a sump screen periodically increases or decreases up to 2000 second, afterwards its peak decreases.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.2038-2045
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• 2021

The main objective of this work is to propose a reliable routine standard operation procedures (SOP) for structural health monitoring and diagnosis of nuclear power plants (NPPs). At present, NPPs have monitoring systems that can be used to obtain the quantitative health record of containment (CTMT) buildings through system identification technology. However, because the measurement signals are often interfered with by noise, the identification results may introduce erroneous conclusions if the measured data is directly adopted. Therefore, this paper recommends the SOP for signal screening and the required identification procedures to identify the dynamic characteristics of the CTMT of NPPs. In the SOP, three recommend methods are proposed including the Recursive Least Squares (RLS), the Observer Kalman Filter Identification/Eigensystem Realization Algorithm (OKID/ERA), and the Frequency Response Function (FRF). The identification results can be verified by comparing the results of different methods. Finally, a preliminary CTMT healthy record can be established based on the limited number of earthquake records. It can be served as the quantitative reference to expedite the restart procedure. If the fundamental frequency of the CTMT drops significantly after the Operating Basis Earthquake and Safe Shutdown Earthquake (OBE/SSE), it means that the restart actions suggested by the regulatory guide should be taken in place immediately.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1669-1680
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• 2019

Seismic safety is considered to be one of the key design objectives of AP1000 nuclear power plant (NPP) in strong earthquakes. Dynamic behavior, damage development and aggravation effect are studied in this study for the three main components of AP1000 NPP, namely reinforced concrete shield building (RCSB), steel vessel containment (SVC) and reinforced concrete auxiliary building (RCAB). Characteristics including nonlinear concrete tension and compressive constitutions with plastic damage are employed to establish the numerical model, which is further validated by existing studies. The author investigates three earthquakes and eight input levels with the maximum magnitude of 2.4 g and the results show that the concrete material of both RCSB and RCAB have suffered serious damage in intense earthquakes. Considering RCAB in the whole NPP, significant damage aggravation effect can be detected, which is mainly concentrated at the upper intersection between RCSB and RCAB. SVC and reinforcing bar demonstrate excellent seismic performance with no obvious damage.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.659-660
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• 2005

• Journal of the Earthquake Engineering Society of Korea
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• v.17 no.2
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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.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.448-456
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• 2016

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies themass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The $H_2O$ content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of $CO_2$ necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.