• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.05a
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• pp.226-227
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• 2018

Recently, as the interest of the government and the public on energy saving has increased, the airtightness of buildings has been improved to improve the insulation performance of buildings. However, indoor air pollution due to increase of pollution source in indoor space and lack of ventilation is increasing and interest in indoor air quality is increasing. In 2003, the Ministry of Environment enacted and promulgated the Act on Indoor Air Quality Control in Multi-use Facilities. Radon is a naturally occurring radioactive inert gas with colorless, tasteless and odorless nature. The concentration is high in a room where radon can not escape. Although lononggas is naturally occurring, it is not interested in living environment, but it is easily inhaled through human body through respiration and causes lung cancer in long-term exposure. Therefore, this study intends to carry out an experiment for the reduction of radon gas, which is the first carcinogen in indoor air pollution sources.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.879-884
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• 1995

The results of performance assessment analyses have shown that the long-lived radionuclides such as I-129 control the potential individual dose impact to the public. I-129 is difficult-to-measure(DTM) in low-level waste because it is non-gamma emitting radionuclides and exists at extremely low concentrations in radioactive waste generated by nuclear reactors. In this study, computer modeling technique to predict release rate of I-129 is developed to provide another tools far performance assessment of land disposal facilities and characteristics of radwaste. Model suggested in this study will give conservative values of I-129 release rate far determination of radwaste characteristics. More detailed approach is implemented to account for release conditions of fuel source-nuclides. 1-131 concentration measured from reactor coolant and released fraction from tramp fuel have dominant roles in calculating release rate of I-129 with fuel defect conditions.

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.683-690
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• 2007

Because of increasing concerns regarding global warming and the longevity of oil and gas reserves, the importance of nuclear energy as a major source of sustainable energy is gaining recognition worldwide. To make nuclear energy truly sustainable, it is necessary to ensure not only the sustainability of the fuel supply but also the sustained availability of waste repositories, especially those for high-level radioactive waste (HLW). From this perspective, the effort to maximize the waste loading density in a given repository is important for easing repository capacity problems. In most cases, the loading of a repository is controlled by the decay heat of the emplaced waste. In this paper, a comparison of the decay heat characteristics of HLW is made among the various fuel cycle options. It is suggested that, for a future fast breeder reactor (FBR) cycle, the removal and burning of minor actinides (MA) would significantly reduce the heat load in waste and would allow for a reduction of repository size by half.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.11-39
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• 2018

Tritium and $^{14}C$ are two most important radionuclides released from nuclear facilities to the environment, and $^{14}C$ contributes dominant radiation dose to the population around nuclear power plants. This paper presents an overview of the production, pathway, species and levels of tritium and $^{14}C$ in nuclear facilities, mainly nuclear power plants. The methods for sampling and collection of different species of tritium and $^{14}C$ in the discharge gas from the stack in the nuclear facilities, atmosphere of the nuclear facilities and environment are presented, and the features of different methods are reviewed. The on-line monitoring methods of gaseous tritium and $^{14}C$ in air and laboratory measurement methods for sensitive determination of tritium and $^{14}C$ in collected samples, water and environmental solid samples are also discussed in detailed. Meanwhile, the challenges in the determination and speciation analysis of tritium and $^{14}C$ are also highlighted.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.368-376
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• 2004

The estimation of the diffusion coefficients of the Gaussian plume model and the release rate by assimilation of tracer-gas measurements on Younggwang site was tested. Diffusion coefficients were modified by linear programming of both the measurements and the simulated using the Gaussian plume model. The application of the modified diffusion coefficients improved the prediction ability of the Gaussian plume model on both 3 km and 8 km arc lines. And, the release rate of tracer gas was estimated using least squares method. The optimal source rate was estimated by minimizing the errors between the measured concentrations and the computed ones by the Gaussian plume model. The obtained release rate showed a good agreement with the real release rate of the Younggwang experiment in 24%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.17-21
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• 2005

Shielded laser ablation system composed of laser system, image analyser, XYZ translator with motion controller, ablation chamber, manipulator and various optics was designed. Nd:YAG laser which can be tunable from 1064 nm to 266 m was selected as light source. CCD camera(< $\pm$200) was chosen to analyze a crater less than 50 un in diameter. XYZ translator was composed of three linear stage which can travel 50 w with a minimum movement of 1 um and motion controller. Before the performance test, each part of system was optically aligned. To perform the ablation test, the specimen was ablated by 50 um interval and observed by image analyser The shape of crater was almost round, indicating laser beam has homogeneous energy distribution. The resolution and magnification of image system were compatible with the design.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.197-201
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• 2012

Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.

• Journal of the Mineralogical Society of Korea
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• v.27 no.4
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• pp.235-242
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• 2014

Uranium can be released into the water environment from natural sources and human activities. The natural source of uranium is dominant in the Korean soil and groundwater environments. Uranium has both of radioactive and chemical toxic properties. Therefore, a drinking water contaminated with uranium has a high health risk. This study was conducted to determine the uranium concentration of water systems including small village drinking water system, groundwater for drinking water purpose, spring water, groundwater monitoring well, and emergency water suppling system. The uranium concentration was compared with domestic and other countries' standard. The contamination level was also evaluated on the basis of geological characteristics of the area. Among total 803 samples, 6 exceeded the Korean standard, $30{\mu}g/{\ell}$ and this was about 0.7% of the total sample. On the basis of geology, uranium concentration appeared to be increased in order of biotite granodiorite > biotite granite > gneissoid granite. The highest level of uranium was 12.4 in average.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.584-589
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• 2003

The hot cell facility which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ and $1.01{\times}10^{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ and $7.88{\times}10^{-2}$ mSv/h, respectively The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources, and penetration and toboggan will be partly reinforced by lead shield.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.153-164
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• 2014

Currently, there are many programs which are now being developed or already developed to predict radionuclide and corrosion product at the stage of designing NPP. However, since there are not many developments in evaluating quantity of activation corrosion products occurring when disassembling a nuclear power plant there exist some difficulties in calculating accurately. In order to evaluate activation products inventory for the research of effect of neutron activation in the reactor vessel, component of nuclear reactor and adjacent structures, it should be evaluated by using operation history of nuclear reactor, material composition of structure and average neutron flux in every field representing fixed structure of nuclear reactor. In this study, CORA, PACTOLE, CRUDSIM, CREAT and ACE codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor which is used at the stage of designing. As a future study, the accuracy in calculating the quantity of product corrosion can be increase by finding out the possibility of use and improvement for evaluation of the decontamination.