• Journal of the Earthquake Engineering Society of Korea
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• v.19 no.1
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• pp.1-11
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• 2015

In order to perform a soil-isolation-structure interaction analysis of seismically isolated nuclear power plant (NPP) structures, the nonlinear behavior of a seismic isolation system may be converted to an equivalent linear model used in frequency domain analysis. Seismic responses for seismically isolated NPP containment structures subjected to a simple artificial acceleration history and different site class earthquakes are evaluated for the equivalent-linear and nonlinear models that have been applied to lead-rubber bearing (LRB) modeling. It can be observed that the maximum displacements of the equivalent linear model are larger than that of the nonlinear model. From the floor response spectrum analysis for the top of NPP containment structures, it can be observed that the spectral acceleration of an equivalent linear model at about 0.5 Hz frequency is about 2~3 times larger than that of a nonlinear model.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.512-527
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• 2004

This paper describes the finite element (FE) analysis results of a 1/4 scale model of a prestressed concrete containment vessel (PCCV) by considering the tension stiffening effect, which is a result of the bond effect between the concrete and the steel. The tension stiffening model is assumed to be an exponential form based on the relationship between the average stress and the average strain of the concrete. The objective of the present FE analysis is to evaluate the ultimate internal pressure capacity of the PCCV, as well as its failure mechanism, when the PCCV model is subjected to a monotonous internal pressure beyond is design pressure capacity. With the commercial code ABAQUS, the FE analysis used two concrete failure criteria: a 2-dimensional axi-symmetric model with modified Drucker-Prager failure criteria and a 3-dimensional model with a damaged plasticity mod디. The results of our FE analysis on the ultimate pressure capacity and failure modes of PCCV have a good agreement with the experimental data.

• Proceedings of the KIEE Conference
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• 2003.11c
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• pp.916-919
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• 2003

In this paper, we deal with a development of measurement system to apply the leakage rates test of primary containment in nuclear power plant. The measurement test about leakage rates in primary containment is one sort of test to prove safety of nuclear power plant. The parameters which are measured to calculate leakage rates are drybulb temperature, dew point temperature(or relative humidity), absolute pressure and flow. Overall, the measurement system consists of sensor module for data acquisition of the parameters, transfer module for wireless data communication and control module to control system and to calculate leakage rates. Because existing measurement systems are difficult to set in field, we pursued convenience of use, we applied wireless data communication and individual form module using battery. We also changed for the better in confidence. Recently, we are developing a drybulb temperature and a dew point temperature sensor module. We describe about function of developed measurement system, its standard and an plan for verification of measurement system.

• The Journal of the Korea Contents Association
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• v.9 no.6
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• pp.25-36
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• 2009

In this paper, we propose an efficient data management scheme for path queries and containment queries which are occurred frequently. The proposed data management scheme considers a change of the containment of products during a transport and supports a path of changed products by representing a path of various containments. Also, the compression utilizing the structure of supply chain reduces the stored data volumes. As a result, our method outperforms the existing methods in terms of storage efficiency and query processing time.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.827-837
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• 2015

Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR), Organization for Economic Co-operation and Development (OECD)-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS) has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model), based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1345-1354
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• 2019

Clogging of sump strainers that filter the recirculation water in containment after a loss-of-coolant accident (LOCA) seriously impedes the continued cooling of nuclear reactor cores. In experiments examining the corrosion of aluminium alloy 6061, a common material in containment equipment, in borated solutions simulating the water chemistry of sump water after a LOCA, we found that Fe-bearing intermetallic particles, which were initially buried in the Al matrix, were progressively exposed as corrosion continued. Their cathodic nature $vis-{\grave{a}}-vis$ the Al matrix provoked continuous trenching around them until they were finally released into the test solution. Such particles released from Al alloy components in a reactor containment after a LOCA will be transported to the sump entrance with the recirculation flow and trapped by the debris bed that typically forms on the strainer surface, potentially aggravating strainer clogging. These Fe-bearing intermetallic particles, many of which had a rod or thin strip-like geometry, were identified to be mainly the cubic phase ${\alpha}_c-Al(Fe,Mn)Si$ with an average size of about $2.15{\mu}m$; 11.5 g of particles with a volume of about $3.2cm^3$ would be released with the dissolution of every 1 kg 6061 aluminium alloy.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1387-1401
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• 2018

An aircraft impact (AI) on a nuclear power plant (NPP) is considered to be a beyond-design-basis event that draws considerable attention in the nuclear field. As some NPPs have already adopted the seismic isolation technology, and there are relevant standards to guide the application of this technology in future NPPs, a new challenge is that nuclear power engineers have to determine a reasonable method for performing AI analysis of base-isolated NPPs. Hence, dynamic influences of the seismic isolation on the vibration and structural damage characteristics of the base-isolated CPR1000 containment are studied under various aircraft loads. Unlike the seismic case, the impact energy of AI is directly impacting on the superstructure. Under the coupled influence of the seismic isolation and the various AI load, the flexible isolation layer weakens the constraint function of the foundation on the superstructure, the results show that the seismic isolation bearings will produce a large horizontal deformation if the AI load is large enough, the acceleration response at the base-mat will also be significantly affected by the different horizontal stiffness of the isolation bearing. These concerns require consideration during the design of the seismic isolation system.

• Journal of the Korean GEO-environmental Society
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• v.12 no.6
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• pp.35-42
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• 2011

In this paper, model experiments in the laboratory were carried out to predict the effluent concentrations of suspended solid in containment of dumping dredged soils and test results were compared with results estimated by the currently used design method. Model tests of simulating dumping the dredged soils with a pump dredger in field were performed with changing the influent concentration and the length of containment and effluent concentration of suspended solid with time were measured during tests. As results of comparing test results about effluent concentration with those estimated from the design method by US Army COE(1987), they were confirmed to be in relatively good agreements.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2488-2498
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• 2021

The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.