• 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.

• Nuclear Engineering and Technology
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• v.19 no.2
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• pp.107-114
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• 1987

Using a solution of $^{137}$ Cs as a tracer, radioactivity of a $^{137}$ Cs solution is measured by coincidence and anticoincidence counting methods. Sample sources in this study are fabricated by mixing $^{137}$ Cs with tracer, and the mixing ratios are 1/3 and 1/5, respectively in coincidence and anticoincidence methods. Specific activities of a $^{137}$ Cs solution are calculated by means of f-channel low energy threshold level variation and efficiency extrapolation. The results obtained by coincidence and anticoincidence counting methods are 552.78 kBq/g and 554.32kBq/g, respectively on the reference date and the combined uncertainties are 1.60kBq/g and 1.51kBq/g for each method. These two results show a good agreement within the uncertainty range.

• Korean Chemical Engineering Research
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• v.59 no.3
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• pp.450-454
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• 2021

Currently, gamma nuclide analysis is mainly used using inorganic scintillators or semiconductor detectors. These detectors have high resolution but there are less economical, limited in size, and low process ability than plastic scintillators. Therefore, quantum dot-based plastic scintillator was developed using the advantages of the quantum dot nanomaterial and the conventional plastic scintillator. In this study, efficient plastic scintillator was fabricated by adding CdS/ZnS based on the most widely used Cd-based nanomaterial in a polystyrene matrix. In addition, the performance of the commercial plastic scintillator was compared and it was analyzed through radiological measurement experiments. The detection efficiency of fabricated plastic scintillator was higher than commercial plastic scintillator, EJ-200. It is believed that this fabricated plastic scintillator can be used as a radioactivity analyzer in the medical and nuclear facility fields.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.84-85
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• 2009

New approaches for detecting, preventing and remedying environmental damage are important for protection of the environment. Procedures must be developed and implemented to reduce the amount of waste produced in chemical processes, to detect the presence and/or concentration of contaminants and decontaminate fouled environments. Contamination can be classified into three general types: airborne, surface and structural. The most dangerous type is airborne contamination, because of the opportunity for inhalation and ingestion. The second most dangerous type is surface contamination. Surface contamination can be transferred to workers by casual contact and if disturbed can easily be made airborne. The decontamination of the surface in the nuclear facilities has been widely studied with particular emphasis on small and large surfaces. The amount of wastes being produced during decommissioning of nuclear facilities is much higher than the total wastes cumulated during operation. And, the process of decommissioning has a strong possibility of personal's exposure and emission to environment of the radioactive contaminants, requiring through monitoring and estimation of radiation and radioactivity. So, it is important to monitor the radioactive contamination level of the nuclear facilities for the determination of the decontamination method, the establishment of the decommissioning planning, and the worker's safety. But it is very difficult to measure the surface contamination of the floor and wall in the highly contaminated facilities. In this study, the poly(styrene-ethyl acrylate) [poly(St-EA)] core-shell composite polymer for measurement of the radioactive contamination was synthesized by the method of emulsion polymerization. The morphology of the poly(St-EA) composite emulsion particle was core-shell structure, with polystyrene (PS)as the core and poly(ethyl acrylate) (PEA) as the shell. Core-shell polymers of styrene (St)/ethyl acrylate (EA) pair were prepared by sequential emulsion polymerization in the presence of sodium dodecyl sulfate (SOS) as an emulsifier using ammonium persulfate (APS) as an initiator. The polymer was made by impregnating organic scintillators, 2,5-diphenyloxazole (PPO) and 1,4-bis[5-phenyl-2-oxazol]benzene (POPOP). Related tests and analysis confirmed the success in synthesis of composite polymer. The products are characterized by IT-IR spectroscopy, TGA that were used, respectively, to show the structure, the thermal stability of the prepared polymer. Two-phase particles with a core-shell structure were obtained in experiments where the estimated glass transition temperature and the morphologies of emulsion particles. Radiation pollution level the detection about under using examined the beta rays. The morphology of the poly(St-EA) composite polymer synthesized by the method of emulsion polymerization was a core-shell structure, as shown in Fig. 1. Core-shell materials consist of a core structural domain covered by a shell domain. Clearly, the entire surface of PS core was covered by PEA. The inner region was a PS core and the outer region was a PEA shell. The particle size distribution showed similar in the range 350-360 nm.

• Korean Journal of Soil Science and Fertilizer
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• v.4 no.1
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• pp.55-66
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• 1971

A method for the identification of and measurement of beta activity due to $^{40}K$ have been developed in this institute. The method is based on the principle of : $$G(t)=\frac{A}{A_{\infty}}=1-e^{-{\eta}t}$$ where: G(t)=fraction of maximum activity A = counting rate of thickness $A_{\infty}$= saturation activity ${\eta}$= mass absorption coefficient of $^{40}K$ By this technique, total beta activity in 92 Korean paddy soil samples collected from various part of the country, have been determined and the results of this analysis reported in this paper. Most of the beta activity in soils have been accounted for to be due to $^{40}K$.