• The Journal of the Korea Contents Association
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• v.14 no.6
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• pp.255-261
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• 2014

The generation amount of radioactive waste has been rapidly increased by increase of the usage of radioisotope source in medical field. Especially, the use of the radioactive source of I-131 with short half-life of 8.02 days used in treatment of thyroid has been increased, and all of the wastes concerned have been disposed by means of the self-disposal method. IAEA proposed criteria for clearance level of waste which depends on the individual (10 ${\mu}Sv/y$) and collective dose (1 man-Sv/y), and concentration of each nuclide (IAEA Safety Series No 111-P-1.1, 1992 and IAEA RS-G-1.7, 2004). In this study, various radioactive wastes in medical fields are collected and measured for establishing the disposal methods and procedures of radioactive wastes. In addition, comparison evaluation of decay storage period between the half-life which was calculated by attenuation curve based on real measurement and analytical half-life is considered. With comparing the theoretical half-life and the effective half life(7.72 days) which was based on the decay equation of measured data, it is resulted in the theoretical half-life is longer than effective half-life. The storage period of radioactive waste for self-disposal may be curtailed. The result of this study will be proposed as ISO standard.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.503-510
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• 2020

Concrete is one of the most widely used materials as the shielding structures of a nuclear facilities. It is also the most generated radioactive waste in quantity while dismantling facilities. Since the concrete captures neutrons and generates various radionuclides, radiation measurement and analysis of the sample was fulfilled prior to dismantle facilities. An HPGe detector is used in general for the radiation measurement, and effective correction factors such as geometrical correction factor, self-absorption correction, and absolute detector efficiency have to be applied to the measured data to decide exact radioactivity of the sample. Correction factors are obtained by measuring data using a standard source with the same geometry and chemical states as the sample under the same measurement conditions. However, it is very difficult to prepare standard concrete sources because concrete is limited in pretreatment due to various constituent materials and high density. In addition, the concrete sample obtained by core drill is a volumetric source, which requires geometric correction for sample diameter and self absorption correction for sample density. Therefore in recent years, many researchers are working on the calculation of effective correction factors using Monte carlo simulation instead of measuring them using a standard source. In this study we calculated, using Geant4, one of the Monte carlo codes, the correction factors for the various diameter and density of the concrete core sample at the gamma ray energy emitted from the nuclides ¹⁵²Eu and ⁶⁰Co, which are the most generated in radioactive concrete.

• Journal of Radiation Protection and Research
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• v.43 no.4
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• pp.160-169
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• 2018

Background: This study aims to develop an integrated optical system that can simultaneously or selectively measure the signals obtained from radioluminescence (RL), thermoluminescence (TL), and optically stimulated luminescence (OSL), which are luminescence phenomena of materials stimulated by radioactivity, heat, and light, respectively. The luminescence mechanism of various materials could be investigated using the glow curves of the luminescence materials. Materials and Methods: RL/TL/OSL integrated measuring system was equipped with a X-ray tube (50 kV, $200{\mu}A$) as an ionizing radiation source to irradiate the sample. The sample substrate was used as a heating source and was also designed to optically stimulate the sample material using various light sources, such as high luminous blue light emitting diode (LED) or laser. The system measured the luminescence intensity versus the amount of irradiation/stimulation on the sample for the purpose of measuring RL, TL and OSL sequentially or by selectively combining them. Optical filters were combined to minimize the interference of the stimulation light in the OSL signal. A long-pass filter (420 nm) was used for 470 nm LED, an ultraviolet-pass filter (260-390 nm) was used for detecting the luminescence of the sample by PM tube. Results and Discussion: The reliability of the system was evaluated using the RL/OSL characteristics of $Al_2O_3:C$ and the RL/TL characteristics of LiF:Mg,Cu,Si, which were used as dosimetry materials. The RL/OSL characteristics of $Al_2O_3:C$ showed relatively linear dose-response characteristics. The glow curve of LiF:Mg,Cu,Si also showed typical RL/OSL characteristics. Conclusion: The reliability of the proposed system was verified by sequentially measuring the RL characteristics of radiation as well as the TL and OSL characteristics by concurrent thermal and optical stimulations. In this study, we developed an integrated measurement system that measures the glow curves of RL/TL/OSL using universal USB-DAQs and the control program.

• Proceedings of the Korean Society of Disaster Information Conference
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• 2015.11a
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• pp.294-297
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• 2015

Decommissioning of nuclear facilities should be accomplished by assuring the safety of workers because decommissioning activities of nuclear facilities are under high radioactivity and work difficulty. It is necessary that before decommissioning, the radiation exposure dose of workers has to be evaluated and assessed under the principle of ALARA (as low as reasonably achievable). Furthermore, to improve the proficiency of decommissioning environments, method and system need to be developed. The legacy methods of exposure dose measurement and assessment had the limitations to modify and simulate the exposure dose to workers prior to practical activities because those should be accomplished without changes of working routes under predetermined scenarios. To simulate a lot of decommissioning scenarios, decommissioning environments were designed in virtual reality. To simulate and assess the exposure dose to workers, human model also was designed in virtual environments. These virtual decommissioning environments made it possible to real-time simulate and assess the exposure dose to workers. It can be concluded that this system is able to protect from accidents and enable workers to improve his familiarization about working environments. It is expected that this system can reduce human errors because workers are able to improve the proficiency of hazardous working environments due to virtual training like real decommissioning situations. In the end, the safety during decommissioning of nuclear facilities will be guaranteed under the principle of ALARA.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.426-427
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• 2005

Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical process in which the spent fuel of PWR is transformed into the uranic metal ingot. Through this process, which has been developed in KAERI since 1998, the radioactivity, the radiotoxicity, the heat and the volume of the PWR spent fuel are reduced by a quarter of the original. To demonstrate a lab-scale process and extract the data for the later pilot-scale process, a demonstration facility of ACP (ACPF) is under construction and the lab-scale demonstration is slated for 2006. To establish the safeguardability of ACPF, a safeguards system including a neutron counter based on non-destructive assay, which is named as ACP Safeguards Neutron Counter (ASNC), the ACP Safeguards Surveillance System (ASSS) which consists of two neutron monitors and five IAEA cameras, and Laser Induced Breakdown System (LIBS) have been developed and are ready to be installed at ACPF. The target materials of ACP to assay with ASNC are categorized into three types among which the first is the uranic metal ingot, the second is the salt waste and the last is $UO_2$ and $U_{3}O_8$ powders, rod cuts and hulls. The Pu content of process nuclear materials can be accounted with ASNC. The ASSS is integrated in the ACP Intelligent Surveillance Software (AISS) in which the IAEA camera images and background signals at the rear doors of ACPF are displayed. The composition of special nuclear materials of ACP can be measured with LIBS which can be a supporting measurement tool for ASNC. The conceptual picture of safeguards system of ACPF is shown in Fig. 1.

• Journal of Biomedical Engineering Research
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• v.5 no.2
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• pp.167-172
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• 1984

In the present study, we developed a nuclear stethoscope interfaced with a personal computer (APPLE- II) and evaluated it's performances as compared with the gamma camera. This single-probe, ECG-gated system has been used for displaying the left ventricular time-activity curves, the peak ejection time and its rate, the peak filling time and its rate, and the R-R interval distribution. The radioactivity was measured at every 10msec around the ventricular region, where the activity curves shows the maximal peak to peak variations. The background activity was measured around the lung area showing its counts approximately 50% of the end-diastolic count with minimal variations. The average time-activity curves of 100 beats were used for analysis in the equilibrium study after intravenous injection of 15-20mCi of Tc. The ejection fractions measured by the nuclear stethoscope(Y) were compared with those measured by gamma camera(X) in 47 patients with various heart diseases. The correlation coefficient between two measurements was 0.766 with a relation of Y=1.04 x-8.48. Also, the high reproducibility was obtained for the same patient. Also, the high reproducibility was obtained for the same patient. From this study, we conclude that this device is useful for continuous monitoring in the intensive care unit, as it is portable, compact, and inexpensive.

• Journal of Advanced Marine Engineering and Technology
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• v.38 no.4
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• pp.470-475
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• 2014

Recently, smart phones are spread out as a universal mobile terminal equipment. Its provide USB interface to connect with various devices. Smart phones may be replaced with high priced monitoring equipments because of portability and mobility as its merit. In addition to that, the various sensor devices detecting surrounding environment of the radioactivity, sodium or electromagnetic waves have been announced. But the plug and play methods of sensor devices have some problem to connect smart phone with USB accessory device. We propose an integrated methodology that can connect smart phone with USB sensor devices and also, we realized USB accessory plug and play with one App that can collect measurement data through various sensor devices.

• Korean Journal of Remote Sensing
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• v.27 no.5
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• pp.573-586
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• 2011

The MTF(modulation transfer function) is one of parameters to evaluate the performance of imaging systems. Also, it can be used to restore information that is lost by a harsh space environment (radioactivity, extreme cold/heat condition and electromagnetic field etc.), atmospheric effects and falloff of system performance etc. This paper evaluated the MTF values of images taken by DubaiSat-1 satellite which was launched in 2009 by EIAST(Emirates Institute for Advanced Science and Technology) and Satrec Initiative. Generally, the MTF was assessed using various methods such as a point source method and a knife-edge method. This paper used the slanted-edge method. The slantededge method is the ISO 12233 standard for the MTF measurement of electronic still-picture cameras. The method is adapted to estimate the MTF values of line-scanning telescopes. After assessing the MTF, we performed the MTF compensation by generating a MTF convolution kernel based on the PSF(point spread function) with image denoising to enhance the image quality.

• Journal of Radiation Protection and Research
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• v.33 no.3
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• pp.121-126
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• 2008

Radioactivities of the stack air effluents from the Advance Fuel Science Building (AFSB) at KAERI have been investigated and evaluated. In this AFSB, nuclear fuels for the HANARO research reactor have been fabricated and the advanced nuclear fuels have been studied. A stack air monitoring system has been continuously operating to monitor the stack air effluents from the facility to protect the environment. As the results of the periodical radioactivity measurement and both the gamma and alpha spectrometry for the millipore filters taken from the stack air monitor from January until March 2008, a trace amount of primordial $^{40}K$ and the short-lived decay products of natural borne $^{222}Rn$ and $^{220}Rn$ have been detected. However, the radioactivities have rapidly decayed to the level below the Minimum Detectable Activity (MDA) of the counting system. Therefore, it was evaluated that no uranium isotopes have been released to the atmosphere from the stack of the AFSB at KAERI.

• Journal of Radiation Protection and Research
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• v.34 no.3
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• pp.121-128
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• 2009

During the maintenance period at Korean nuclear power plants, internal exposure of radiation workers occurred by the inhalation of $^{131}I$ released to the reactor building when primary system was opened. The internal radioactivity of radiation workers contaminated by $^{131}I$ was immediately measured using a whole body counter and the whole body counting was performed again after a few days. In this study, the intake estimated from the record history of entrance to radiation control areas and the measurement results of air sampling for $^{131}I$ in those areas, were compared with that from the results of whole body counting. As a result, it was concluded that the intake estimation using whole body counting and air sampling showed similar results.