• International Journal of Contents
• /
• v.8 no.2
• /
• pp.60-66
• /
• 2012

This study was designed to detect and measure the concentration of radioactivity of natural radionuclides ($^{238}U$, $^{232}Th$, $^{40}K$) and artificial radionuclides ($^{137}Cs$, $^{60}Co$) present in the drinking water of the city of Busan and surrounding areas in South Korea, and also to measure the absorbed dose of radiation caused by these elements in the residents so as to help better manage the risk that these radionuclides pose in the future. For the purposes of the study, a total of 42 samples of water were collected from three key water sources (19 samples of groundwater, 4 samples of tap water, and 19 samples of surface water) and their contents were analyzed for radioactivity concentration. The results revealed that two natural radionuclides, $^{238}U$ and $^{232}Th$, exist in the groundwater with an average concentration of radioactivity of 3.34 Bq/L and $8.28{\times}10^{-5}Bq/L$ respectively, while the surface water was found to contain the same two radionuclides with mean concentrations of 0.849 Bq/L and $1.103{\times}10^{-4}Bq/L$ respectively. In addition, of the 19 samples of the groundwater, $^{137}Cs$ was found in eight of them and $^{60}Co$ was detected in ten. Of the four samples of the tap water, $^{137}Cs$ was detected in all samples and $^{60}Co$ was detected in three. Both $^{137}Cs$ and $^{60}Co$ were detected in all 12 samples of surface water. As far as $^{40}K$ is concerned, this element was detected in three of the 19 groundwater samples, but was not detected in any surface or tap water sample. In addition, the absorbed dose of $^{238}U$ from the groundwater was $7.94{\times}10^{-8}Sv/y$, while the absorbed dose of $^{232}Th$ from the surface water was $9.33{\times}10^{-13}Sv/y$. The absorbed dose of $^{137}Cs$ from the tap water was $7.33{\times}10^{-5}Sv/y$, while the absorbed dose of $^{60}Co$ from the surface water was the highest at $4.23{\times}10^{-6}Sv/y$.

• Analytical Science and Technology
• /
• v.12 no.5
• /
• pp.428-435
• /
• 1999

A method for measuring Rn-222 and Ra-226 in aqueous sample using liquid scintillation counting technique has been studied. The Rn-222 was extracted easily from the water sample (10 mL) by 12 mL of xylene based organic scintillant. After radioactive equilibrium between Rn-222 and its alpha emitting decay products for three hours, the alpha activity from Rn-222 and its decay products were measured in a scintillation vial using the Wallae $1220^{TM}$ Quantulus liquid scintillation counter. Ra-226 concentration in aqueous sample was determined, after isolation of Ra-226 from the sample matrix, by extraction the ingrowth of the Rn-222 and its alpha emitting decay products with xylene based organic scintillant. The optimum pulse-shape analysis (PSA) value was evaluated by the figure of merit (FM) criterion. Minimum detectable activity (MDA) is about 0.14 Bq/L (3.78 pCi) for the region of Rn-222 and its alpha emitting decay products and 0.06 Bq/L (1.63 pCi) for the region of Po-214 respectively, with 200 min, counting time at PSA level 100 in the low-diffusion polyethylene vial and xylene based cocktail solution. Experiment on the optimum sample-cocktail volume ratio, the influence of agitation and the diffusion of radon from vial were carried out.

• Korean Journal of Food Science and Technology
• /
• v.47 no.6
• /
• pp.789-792
• /
• 2015

Samples of fishery products were tested for radioactivity by using the intake frequency data from Korea Health Statistics. The radioactivity of $^{40}K$, $^{137}Cs$, $^{134}Cs$, and $^{131}I$ was analyzed using gamma spectrometry with a simplified sample pre-treatment procedure. The radioactivity range for $^{40}K$ was 21.9-3050 Bq/kg, whereas the radioactivities of $^{137}Cs$, $^{134}Cs$, and $^{131}I$ were under minimum detectable activity which were in the range of 0.140-1.97, 0.0900-1.89 and 0.124-1.94 Bq/kg, respectively, for the three species. The results suggest that the Fukushima accident did not have a significant impact on domestic fishery products, which were analyzed during the period from 2013 to 2015. Additionally, there seemed to be no significant impact of additional exposure dose by the analyzed radionuclides.

• Journal of Radiation Protection and Research
• /
• v.21 no.2
• /
• pp.139-143
• /
• 1996

A multi-wire proportional counter with large sensitive area was designed and constructed considering diameter of anode wire. its material and space. A preamplifier connecting detector to main amplifier or counter was also designed and constructed for measurement output pulse from multi-wire proportional counter. The preamplifier was composed of charge-sensitive differential circuit. clipping circuit and amplification circuit. To test the performance of this equipment, terminal output pulse from the preamplifier was measured and compared with noise For these tests $^{239}Pu(360 Bq)\;and\; ^{90}Sr/^{90}Y(250 Bq)$ were used as radiation sources. The noise ingredient contributing to the maximum amplitude(180mV from $^{239}Pu$ and 200 mV from $^{90}Sr/^{90}Y$) was found to be very small(8 mV) Piled up pulse occurring at the output pulse of charge-sensitive differential circuit was measured as an independent pulse since this affected the amplification in the clipping circuit and amplification circuit. This information can be used to improve the loss of measurement due to piled up pulse.

• Journal of Radiation Protection and Research
• /
• v.33 no.3
• /
• pp.113-120
• /
• 2008

In this study, The TRITGO code was introduced, which can predict the amount of tritium production, it's transport, removal, distribution and the level of contamination for the produced hydrogen by the tritium on the VHTR (very high temperature gas cooled reactor). The TRITGO code was improved so that the permeation to the IS Iodine Sulfide) loop for producing the hydrogen can be simulated. The contamination level of the produced hydrogen by the tritium was predicted by the improved code for the VHTR with 600MW thermal power. The contamination level for the produced hydrogen by tritium was predicted as 0.055 Bq/$H_2-g$. This level is three order of lower than the regulation value of 56 Bq/$H_2-g$ from Japan. From this study, the following results were obtained. it is important that the fuel coating (SiC layer) should be kept intact to prevent the tritium from releasing. Also it is necessary that the level of impurity such as 3He and Li in the helium coolant and the reflector consisting of the graphite should be kept as low as possible. It was found that the capacity of the purification system for filtering the impurities directly from the coolant will be the important design parameter.

• Analytical Science and Technology
• /
• v.9 no.3
• /
• pp.262-269
• /
• 1996

The muiltiwire proportional counter for the measurement of low-level and environmental $\alpha$ particles emitting nuclides was developed. External dimension of the devloped multiwire proportional counter is $350{\times}290{\times}30mm$ and the sensitivity area is $250{\times}200mm$. The wall material of the detector was selected the stainless steel to prevent the deformation by external impact and to obtain minimum background. The anode and cathode wires were used the stainless steel material of diameter $50{\mu}m$. The spacing of each wires are 10.0mm, 5.0mm and the numbers of total wire are 21, 42 lines, respectively. The multiwire proportional counter was designed that the measurement source is placed within the detector to prevent the wall absorption effect and the efficiency variation by various source heights. The characteristics of the developed detector have been investigated to obtain the plateau, operating voltage, background, counting efficiency, position sensitivity and energy resolution etc. For the $^{241}Am$ nuclide, the calculated LLD(Lower Limit of Detection) is 5.0mBq/L which is lower than 40mBq/L of recommended LLD value by ISO(International Organization for Standardization).

• The Korean Journal of Nuclear Medicine Technology
• /
• v.18 no.2
• /
• pp.28-32
• /
• 2014

Purpose Patient motion during myocardial perfusion SPECT can produce images that show visual artifacts and perfusion defects. This artifacts and defects remain a significant source of unsatisfactory myocardial perfusion SPECT. Motion correction has been developed as a way to correct and detect the patient motion for reducing artifacts and defects, and each motion correction uses different algorithm. We corrected simulated motion patterns with several motion correction methods and compared those images. Materials and Methods Phantom study was performed. The anthropomorphic torso phantom was made with equal counts from patient's body and simulated defect was added in myocardium phantom for to observe the change in defect. Vertical motion was intentionally generated by moving phantom downward in a returning pattern and in a non-returning pattern throughout the acquisition. In addition, Lateral motion was generated by moving phantom upward in a returning pattern and in a non-returning pattern. The simulated motion patterns were detected and corrected similarly to no-motion pattern image and QPS score, after Motion Detection and Correction Method (MDC), stasis, Hopkins method were applied. Results In phantom study, Changes of perfusion defect were shown in the anterior wall by the simulated phantom motions, and inferior wall's defect was found in some situations. The changes derived from motion were corrected by motion correction methods, but Hopkins and Stasis method showed visual artifact, and this visual artifact did not affect to perfusion score. Conclusion It was confirmed that motion correction method is possible to reduce the motion artifact and artifactual perfusion defect, through the apply on the phantom tests. Motion Detection and Correction Method (MDC) performed better than other method with polar map image and perfusion score result.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.1
• /
• pp.25-31
• /
• 2009

An analytical method of $^{99}Tc$ concentration in soil was set up and discussed considering the $^{99}Tc$ concentration in Korean soil measured with its analytical method. A selective TEVA resin was used to separate and purify the $^{99}Tc$ in the soil sample. $^{99m}Tc$ from a commercial $^{99}Mo/^{99m}Tc$ generator was used as a yield tracer for the chemical separation of $^{99}Tc$ and its problem when using $^{99m}Tc$ as a tracer was discussed. The chemical recovery yield of $^{99}Tc$ was above 70%. The optimum conditions of inductively coupled plasma mass spectrometry system(ICP-MS) were set up to determine the $^{99}Tc$ after the separation process. The minimum detectable activity(MDA) was 15 mBq/kg-dry in this analytical procedure. The $^{99}Tc$ concentration in soils of Jeju and Kori were measured in the rage of 33.73-89.16 mBq/kg-dry. Those values were less than those reported in other countries and seemed to be originated from atmospheric fallout.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.3
• /
• pp.193-200
• /
• 2005

[ $^{14}C$ ] and $^{3}H$ in the evaporator bottom (EB) discharged from the Nuclear power plant (NPP) were extracted simultaneously into a gaseous $^{14}CO_{2}$ and liquefied HTO by using the chemical oxidation, which is the method to oxidize samples completely using potassium persulfate and sulfuric acid. The extracted $^{14}C$ and $^{3}H$ were counted by the liquid scintillation counter (LSC) after the quench correction. To examine the recovery of $^{14}C$ using the radioactive standards, $Na_{2}^{14}CO_{3}$, $^{14}C-alcohol$, and $^{14}C-toluene$ as $^{14}C$, and HTO as $^{3}H$ were used. Also, the most suitable method for oxidizing $^{14}C-toluene$, which is difficult to be oxidized, was investigated through FT-IR spectra according to the concentration of sulfuric acid. With the identical method, $^{14}C$ and $^{3}H$ in the EB generated in the NPP were assayed in the range of $8.35{\sim}l.38{\times}10^3$ Bq/g and $2.46{\times}10^2{\sim}1.40{\times}10^4$ Bq/g, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.2
• /
• pp.95-102
• /
• 2006

Alpha spectrometry was studied for the determination of $^{237}Np$ in spent nuclear fuel samples. The optimum condition for the electrodeposition of $^{237}Np$ was obtained as follows : for $1{\sim}1.5$ hour of deposition time, at the current intensity of $1.2{\sim}1.5$ A and at sodium sulfate electrolyte without organic additive. The deposition yield and its reproducibility on $^{237}Np$ was decreased as the amount of $^{237}Np$ decreased from 4.16 Bq down to 0.0264 Bq(1ng). The recovery yield of $^{237}Np$ determined by alpha spectrometry after separation in synthetic solution was $98.8{\pm}5.1%$(n=4). The contents of $^{237}Np$ in spent nuclear fuel samples were determined and the result showed an agreement within 10% of a difference between the measurement and the calculation.