• 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.

• Journal of Radiation Protection and Research
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• v.29 no.4
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• pp.257-261
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• 2004

A coded-aperture gamma camera was developed to increase the sensitivity of a pin hole camera made with a pixellated CsI(Tl) scintillator and a position-sensitive photomultiplier tube. The modified round-hole uniformly redundant array of pixel size $13{\times}11$ was chosen as a coded mask considering the detector spatial resolution. The performance of the coded-aperture camera was compared with the pin hole camera using various forms of Tc-99m source to see the improvement of signal-to-noise ratio or the improvement of the sensitivity. The image quality is much improved despite of a slight degradation of the spatial resolution. Though the camera and the test were made for low energy case, but the concept of the coded-aperture gamma camera could be effectively used for the radioactive environmental monitoring and other applications.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Journal of Radiation Protection and Research
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• v.37 no.3
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• pp.129-137
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• 2012

In the 2007 recommendation, the ICRP evolves from the previous process-based system of practices and intervention to the system based on the characteristics of radiation exposure situation. In addition, ICRP recommends the application of source-related dose constraints under the planned exposure situation as a tool for the optimization of protection to workers and the member of public. In this study, the analysis of radioactive effluents from Korean nuclear power plants and the public dose assessment were conducted in reference with the use of dose constraints. Finally, the measure to implement the dose constraints for the member of public was suggested taking into account multi-unit reactors operating at a single site in Korea.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.155-159
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• 1985

In general, dose rates for a monoenergetic gamma emitter uniformly distributed in an infinite cloud have been calulated by using the monoenergetic point-isotorpic source kernel technique. The most serious limitation on use of the kernel technique is subjected to the fact that it estimates the dose only at the surface of body. As a result, an alternative method is presented in which estimates of dose rate for immersion in a radioactive cloud are resulted from the scattered photon spectra incident on the surface of body. The results are in excellent agreement with other's. Work is currently in progress to apply these results to immersion dose problems associated with absorbed dose distribution in the MIRD phatom.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.111-117
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• 2008

The new type phoswich detector consisting of the ZnS(Ag) and plastic scintillator for alpha/beta-ray simultaneous counting was developed for monitoring radiological contamination inside pipes. The detection performance was estimated using the PSD (pulse shape discrimination) method as a function of distance between the scintillator and radioactive source. The attenuation of particles traveling through a thin film for preventing the detector from being contaminated was experimentally estimated. It is concluded from our investigation that the phoswich detector developed can provide a sufficient alpha/beta-ray discrimination. The application of a thin film for preventing the detector from being contaminated was proven to be feasible.