• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.327-335
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• 1995

The normal shutdown and maintenance period of a nuclear power plant can be remarkably shortened when the examination and maintenance works in steam generator tubes are simultaneously carried out with refueling job. There are nozzle dams to Hock the coolant How from reactor to steam generator. Workers are reluctant to install nozzle dam because of the high radiation exposure and the limited working space in steam generator. Moreover, the heavy weight of present nozzle dam makes it installation and removal works much difficult. In this paper, a lighter KAERI nozzle dam with increased flexural rigidity-to-weight was designed and manufactured by changing the structure design of the present nozzle dam and by selecting new material, carbon fiber-reinforced plastic.

• Journal of Korean Society of Coastal and Ocean Engineers
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• v.28 no.1
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• pp.7-12
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• 2016

In this study, a simplified collision model of pile protective structures against a navigation vessel was proposed and verified. The model of pile protective structure were composed by two plastic hinges at below of cap slab and the inside of ground. A nonlinear equation of motions was developed in consideration of the kinematic energy, potential energy and deformation energy in collision event. The developed simplified model were verified by the precise finite element collision analysis of the vessel and the protective structure.

• Nuclear Engineering and Technology
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• v.36 no.1
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• pp.36-46
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• 2004

The ISAAC fission product release calculation is based on detailed FPRAT models developed by Jaycor. For volatile fission product release calculations, either the Cubicciotti steam oxidation correlation or the NUREG-0772 correlation is used. In this study, evaluation is carried out for these volatile fission product release models. As a result, in the case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, the NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option is evaluated to show mitigated conservative results. In addition, a sensitivity study on detailed core nodalization is performed. In the study, 380 horizontal fuel channels in the Wolsong plant are nodalized into 12 (6 channels per loop, $3{\times}3$ Core Pass) representative channels and detailed by 16/20/24 channels. For reference accidents, LOAH and large LOCA are selected as representing high and low pressure sequences, respectively. According to the results, the original 12 channel approach with $3{\times}3$ core passes is evaluated to be sufficient as an optimal scheme.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.98-113
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• 2016

A reactor vessel level monitoring system measures the water level in a reactor during normal operation and abnormal conditions. A drop in the water level can expose nuclear fuel, which may lead to fuel meltdown and radiation spread in accident conditions. A level monitoring system mainly consists of a sensing line and pressure transmitter. Over a period of time boron sediments or other impurities can clog the line which may degrade the accuracy of the monitoring system. The aim of this study is to determine blockage in a sensing line using the energy of the composite signal. An equivalent Pi circuit model is used to simulate blockages in the sensing line and the system's response is examined under different blockage levels. Composite signals obtained from the model and plant's unblocked and blocked channels are decomposed into six levels of details and approximations using a wavelet filter bank. The percentage of energy is calculated at each level for approximations. It is observed that the percentage of energy reduces as the blockage level in the sensing line increases. The results of the model and operational data are well correlated. Thus, in our opinion variation in the energy levels of approximations can be used as an index to determine the presence and degree of blockage in a sensing line.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.543-550
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• 2012

The decrease in the fracture toughness of ferritic steels in a reactor pressure vessel is an important factor in determining the lifetime of a nuclear power plant. A surveillance program has been in place in Korea since 1979 to assess the structural integrity of RPV steels. In this work, the surveillance data were collected and analyzed statistically in order to derive the empirical relationship between the embrittlement and strengthening of irradiated reactor pressure vessel steels. There was a linear relationship between the yield strength change and the transition temperature shift change at 41 J due to irradiation. The proportional coefficient was about $0.5^{\circ}C$/MPa in the base metals (plate/forgings). The upper shelf energy decrease ratio was non-linearly proportional to the yield strength change, and most of the data lay along the trend curve of the US results. The transition regime temperature interval, ${\Delta}T_T$, was less than the US data. The overall change from irradiation was very similar to the US results. It is expected that the results of this study will be applied to basic research on the multiscale modeling of the irradiation embrittlement of RPV materials in Korea.

• Journal of the Korean Professional Engineers Association
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• v.44 no.6
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• pp.40-44
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• 2011

Nuclear energy in Korea began in 1958, when the Korea's atomic energy act was formulated and the relevant organizations were founded. Since then, notwithstanding the two catastrophe like TMI and Chernobyl accident, Korea made a wise decision to expand the peaceful uses of the nuclear energy as well as to localize the essential nuclear design technology of fuel and nuclear steam supply system. This decision resulted in the success of export of nuclear power plants as well as research reactor in 2010s. The Korea's nuclear policy, which well utilized 'international crisis in nuclear business' as 'opportunity of Korea to get. nuclear technology', is believed nice policy as a role model of nuclear new-comer countries. Based upon the success story of localization of nuclear technology, Korea had an eye for a niche market, which was a basis of development of SMART, Korea-made integral PWR. The operation of a SMART plant can sufficiently provide not only electricity but also fresh water for 100,000 residents. Last two years, Korea's nuclear industry team led by the Korea Atomic Energy Research Institute completed the standard design of SMART and applied to the Korea's regulatory body for standard design approval. Now the Korea's licensing authority is reviewing the design with the relevant documents, and the design team is doing its best to realize its hope to get the approval by the end of this year. From next year, the SMART business including construction and export will be explored by the KEPCO consortium.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.673-682
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• 2018

One of the most widely used methods to estimate core damage during a nuclear power plant accident is containment radiation measurement. The evolution of severe accidents is extremely complex, leading to uncertainty in the containment dose rate (CDR). Therefore, it is difficult to accurately determine core damage. This study proposes to conduct uncertainty analysis of CDR for core damage assessment. First, based on source term estimation, the Monte Carlo (MC) and point-kernel integration methods were used to estimate the probability density function of the CDR under different extents of core damage in accident scenarios with late containment failure. Second, the results were verified by comparing the results of both methods. The point-kernel integration method results were more dispersed than the MC results, and the MC method was used for both quantitative and qualitative analyses. Quantitative analysis indicated a linear relationship, rather than the expected proportional relationship, between the CDR and core damage fraction. The CDR distribution obeyed a logarithmic normal distribution in accidents with a small break in containment, but not in accidents with a large break in containment. A possible application of our analysis is a real-time core damage estimation program based on the CDR.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.683-689
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• 2018

To evaluate equipment survivability of the polymer Viton, used in sealing materials, the effects of its thermal degradation were investigated in severe accident (SA) environment in a nuclear power plant. Viton specimens were prepared and thermally degraded at different SA temperature profiles. Changes in mechanical properties at different temperature profiles in different SA states were investigated. The thermal lag analysis was performed at calculated convective heat transfer conditions to predict the exposure temperature of the polymer inside the safety-related equipment. The polymer that was thermally degraded at postaccident states exhibited the highest change in its mechanical properties, such as tensile strength and elongation.

• Journal of Welding and Joining
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• v.7 no.3
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• pp.28-35
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• 1989

It was reported that the toughness for welded region was influenced by various factors such as the gradient for prior austenite grain size, the variation of microhardness and the characteristic microstructure depending on distance from the fusion boundary. Therefore, in order to evaluate the fracture strength of the weldment in which the microstructures change continuously, it is important to assess the peculiar strength of each microstructure in welded region. It was known that the small punch(SP) test technique which was originally developed to study the irradiation damage effect for the structures of nuclear power plant was also useful to investigate the strength evaluating of nonhomogeneous materials. In this paper, by means of a small punch test technique the possibility of evaluating strength of parent and welded region in SS41 and SM53B steels was investigated. The obtained results are summerized as follows: 1) The small punch test which showed markedly the ductile-brittle transition behavior in this experiment may be applied to evaluation for the fracture strength of welded region. 2) It was shown that the ductile-brittle regime lied in Region III(plastic membrane stretching region) of the flow characteristics observed in SP test. 3) The SP test technique which shows a more precipitous energy change transition behavior than the other test technique is able to estimate the more precise transition temperature. 4) It could be seen that in comparision with the structure of parent the structure of weld HAZ in SS41 steel was improved while it in SM53B steel was deteriorated.

• Journal of Korean Society on Water Environment
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• v.32 no.2
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• pp.191-196
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• 2016

This study aimed to evaluate SBR operation cycle for removing the high-concentration organic matter of distillery wastewater in the ginseng processing plant. The experiment was conducted with the use of a laboratory scale SBR reactor and distillery wastewater as the influent. The results indicated an increase in pH from 4.08 to 7.59 of distillery wastewater after aeration for 2 hours. Also, the optimum SBR operation cycle for the removal of organic matter and nitrogen was 2 hr of aeration and 6 hr of anaerobic conditions. Adjustment of proper pH through aeration time is most critical in the SBR operation for distillery wastewater treatment. In this study, we presented an efficient method for distillery wastewater treatment.