• Journal of Radiation Protection and Research
• /
• v.40 no.3
• /
• pp.124-131
• /
• 2015

Radiation exposure to humans can be caused by the gamma rays emitted from natural radioactive elements(such as uranium, thorium and potassium and any of their decay products) of Naturally Occurring Radioactive Materials(NORM) or Technologically Enhanced Naturally Occurring Radioactive Materials(TENORM) added consumer products. In this study, assume that activity of radioactive elements is $^{238}U$, $^{235}U$, $^{232}Th$ $1Bq{\cdot}g^{-1}$, $^{40}K$ $10Bq{\cdot}g^{-1}$ and the gamma rays emitted from these natural radioactive elements radioactive equilibrium state. In this study, reflected End-User circumstances and evaluated annual exposure dose for products based on ICRP reference voxel phantoms and ICRP Recommendation 103 using the Monte Carlo Method. The consumer products classified according to the adhere to the skin(bracelet, necklace, belt-wrist, belt-ankle, belt-knee, moxa stone) or not(gypsum board, anion wallpaper, anion paint), and Geometric Modeling was reflected in Republic of Korea "Residential Living Trend-distributions and Design Guidelines For Common Types of Household.", was designed the Room model($3m{\times}4m{\times}2.8m$, a closed room, conservatively) and the ICRP reference phantom's 3D segmentation and modeling. The end-user's usage time assume that "Development and Application of Korean Exposure Factors." or conservatively 24 hours; in case of unknown. In this study, the results of the effective dose were 0.00003 ~ 0.47636 mSv per year and were confirmed the meaning of necessary for geometric modeling to ICRP reference phantoms through the equivalent dose rate of belt products.

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

• Analytical Science and Technology
• /
• v.18 no.1
• /
• pp.5-12
• /
• 2005

It has been established that soft error of high precision electronic circuits can be induced by alpha particles emitted from the naturally occurring radioactive impurities such as U, and Th. As the electronic circuits have recently become lower dimension and higher density, these alpha-particle emitting radioactive impurities have to be strictly controlled. The aim of this study is to develop of NAA (Neutron Activation Analysis) and gamma-spectrometry to improve the analytical sensitivity and precision of U and Th. A new NAA method has been established using the HTS (Hydrulic transfer system) irradiation facility which has been used to produce radioisotopes for industries and medicines instead of the PTS (pneumatic transfer system) irradiation facility which has been used in general NAA. When the ultratrace impurities have to be analyzed by NAA, background gamma-ray spectra induced from $^{222}Rn$ and its progenies in air is serious problem. This unstable background has been eliminated or stabilized by the use of a nitrogen purging system. Ultra trace amounts of U (0.1 ng/g) and Th (0.01 ng/g) in high purity silica used for EMC could be analyzed by the use of HTS-NAA and low background gamma-spectrometry.

• Economic and Environmental Geology
• /
• v.42 no.4
• /
• pp.335-342
• /
• 2009

The effects of bacterial lysis on the rate of quartz dissolution were investigated under pH 7 condition using Bacillus subtilis cells which were either irradiated or non-irradiated with gamma ray. The amount of dissolved organic carbon (DOC) which resulted from bacterial lysis increased in slurries of quartz and bacteria mixture over experimental period. Lysis of non-irradiated bacteria led to the elevated concentration of dissolved silicon when compared with abiotic control. Concomitant increase in the amounts of DOC and dissolved silicon over time indicated that lixiviation of silicon from quartz was due to bacterial lysis. Higher amounts of DOC and dissolved silicon were present in the irradiated bacterial slurries than those of non-irradiated bacteria. The enhancement of quartz dissolution in the irradiated bacterial slurries was likely attributed to disruption of organic molecules in the bacterial cells by gamma ray and formation of effective ligands for quartz dissolution. The results suggest that the effects of bacterial lysis on mineral weathering rate should be considered for prediction of time for released radionuclides to migrate to surface biosphere in high level radioactive waste disposal site.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.18 no.1
• /
• pp.33-42
• /
• 2014

Purpose: I-131 scan using High Energy (HE) collimator is generally used. While, Medium Energy (ME) collimator is not suggested to use in result of an excessive septal penetration effects, it is used to improve the sensitivities of count rate on lower dose of I-131. This research aims to evaluate I-131 SPECT/CT image quality using by HE and ME collimator and also find out the possibility of ME collimator clinical application. Materials and Methods: ME and HE collimator are substituted as Siemens symbia T16 SPECT/CT, using I-131 point source and NEMA NU-2 IQ phantom. Single Energy Window (SEW) and Triple Energy Windows (TEW) are applied for image acquisition and images with CTAC and Scatter correction application or not, applied different number of iteration and sub set are reconstructed by IR method, flash 3D. By analysis of acquired image, the comparison on sensitivities, contrast, noise and aspect ratio of two collimators are able to be evaluated. Results: ME Collimator is ahead of HE collimator in terms of sensitivity (ME collimator: 188.18 cps/MBq, HE collimator: 46.31 cps/MBq). For contrast, reconstruction image used by HE collimator with TEW, 16 subset 8 iteration applied CTAC is shown the highest contrast (TCQI=190.64). In same condition, ME collimator has lower contrast than HE collimator (TCQI=66.05). The lowest aspect ratio for ME collimator and HE collimator are 1.065 with SEW, CTAC (+) and 1.024 with TEW, CTAC (+) respectively. Conclusion: Selecting a proper collimator is important factor for image quality. This research finding tells that HE collimator, which is generally used for I-131 scan emitted high energy ${\gamma}$-ray is the most recommendable collimator for image quality. However, ME collimator is also applicable in condition of lower dose, lower sensitive if utilizing energy window, matrix size, IR parameter, CTAC and scatter correction appropriately.

• Journal of Radiation Protection and Research
• /
• v.11 no.1
• /
• pp.57-64
• /
• 1986

This paper described an improved technical method required for proper evaluation of personnel exposures by means of the photographic dosimeter developed by KAERI in lower gamma or X-ray energy regions, with which response of the dosimeter varies significantly. With calibration of the dosimeter in the energy range from 30 to 300 keV, the beam spectrum was carefully selected and specified it adequately. The absorber combinations and absorber thickness used to obtain the specified X-ray spectra from a constant potential X-ray machine were determined theoretically and also experimentally. A correlation between the density and exposure for the four separate energies, such as $49\;keV_{eff},\;154\;keV_{eff}\;250\;keV_{eff}\;and\;662\;keV$, is experimentally determined. As a result, it can be directly evaluated the exposure from the measured response of dosimeter.

• The Korean Journal of Nuclear Medicine
• /
• v.33 no.2
• /
• pp.163-171
• /
• 1999

Purpose: Liquid beta emitter filled in angioplasty balloon could be used to perform endovascular balloon brachytherapy to prevent coronary artery restenosis. We investigated the dosimetry for Re-188-DTPA liquid-filled balloon and medical internal radiation dosimetry in case of balloon leakage. Materials and Methods: We estimated radiation dose from an angioplasty balloon (20 mm length, 3 mm diameter cylinder) to the adjacent vessel wall using Monte Carlo EGS4 code. We obtained time-activity curves of kidneys in normal dog and calculated $T_{max},\;T_{1/2}$. Using MIRDOSE3 program, we estimated absorbed doses to the major organs (kidneys, bladder) and the whole body when we assumed that balloon leaked all the isotope contained. Results: The radiation dose was 17.5 Gy at the balloon surface when we applied 3,700 MBq/ml of Re-188 for 100 seconds, Fifty percent of the energy deposited within 1 mm from the balloon surface. The estimated internal dose to the whole body was 0.005 mGy/MBq and 18.5 mGy for the spillage of 3,700 MBq of Re-188. Conclusion: We suggest that Re-188-DTPA can be used for endovascular balloon brachytherapy to inhibit coronary artery restenosis after angioplasty with tolerable whole body radiation dose in case of balloon rupture.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.4
• /
• pp.343-356
• /
• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.

• The Korean Journal of Nuclear Medicine
• /
• v.33 no.6
• /
• pp.527-536
• /
• 1999

Purpose: Radiotracers that bind to the central benzodiazepine receptor are useful for the investigation of various neurological and psychiatric diseases. [C-11]Flumazenil, a benzodiazepine antagonist, is the most widely used radioligand for central benzodiazepine receptor imaging by PET. We synthesized 3-(2-[F-18]fluoro)flumazenil, a new fluorine-18 ($t_{1/2}$= 110 min) labeled analogue of benzodiazepine receptor imaging agent, and evaluated in vivo for biodistribution in mice. Materials and Methods: Flumazenil (Ro 15-1788) was synthesized by a modification of the reported method. Precursor of 3-(2-[F-18]fluoro)flumazenil, the tosylated flumazenil derivative was prepared by the tosylation of the ethyl ester by ditosylethane. [F-18] labeling of tosyl substitued flumazenil precursor was performed by adding F-18 ion at $85^{\circ}C$ in the hot ceil for 20 min. The reaction mixture was trapped by C18 cartridge, washed with 10% ethanol, and eluted by 40% ethanol. Bidistribution in mice was determined after intravenous injection. Results: The total chemical yield of tosylated flumazenil derivative was ${\sim}40%$. The efficiency of labeling 3-(2-[F-18]fluoro)flumazenil was 66% with a total synthesis time of 50 min. Brain uptakes of 3-(2-[F-18]fluoro)flumazenil at 10, 30, 60 min after injection, were $2.5{\pm}0.37,\;2.2{\pm}0.26,\;2.1{\pm}0.11$ and blood activities were $3.7{\pm}0.43,\;3.3{\pm}0.07,\;3.3{\pm}0.09%ID/g$, respectively. Conclusion: We synthesized a tosylated flumazenil derivative which was successfully labeled with no-carrier-added F-18 by nucleophilic substitution.

• Journal of radiological science and technology
• /
• v.38 no.1
• /
• pp.23-30
• /
• 2015

Positron emission tomography scan has been growing diagnostic equipment in the development of medical imaging system. Compare to 99mTc emitting 140 keV, Positron emission radionuclide emits 511 keV gamma rays. Because of this high energy, it needs to reduce radioactive emitting from patients for radio technologist. We searched the external dose rates by changing distance from patients and measure the external dose rates when we used shielder investigate change external dose rates. In this study, the external dose distribution were analyzed in order to help managing radiation protection of radio technologists. Ten patients were searched (mean age: $47.7{\pm}6.6$, mean height: $165.5{\pm}3.8cm$, mean weight: $65.9{\pm}1.4kg$). Radiation was measured on the location of head, chest, abdomen, knees and toes at the distance of 10, 50, 100, 150, and 200 cm, respectively. Then, all the procedure was given with a portable radiation shielding on the location of head, chest, and abdomen at the distance of 100, 150, and 200 cm and transmittance was calculated. In 10 cm, head ($105.40{\mu}Sv/h$) was the highest and foot($15.85{\mu}Sv/h$) was the lowest. In 200 cm, head, chest, and abdomen showed similar. On head, the measured dose rates were $9.56{\mu}Sv/h$, $5.23{\mu}Sv/h$, and $3.40{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.24{\mu}Sv/h$, $1.67{\mu}Sv/h$, and $1.27{\mu}Sv/h$ in 100, 150, and 200 cm on head. On chest, the measured dose rates were $8.54{\mu}Sv/h$, $4.90{\mu}Sv/h$, $3.44{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.27{\mu}Sv/h$, $1.34{\mu}Sv/h$, and $1.13{\mu}Sv/h$ in 100, 150, and 200 cm on chest. On abdomen, the measured dose rates were $9.83{\mu}Sv/h$, $5.15{\mu}Sv/h$, and $3.18{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.60{\mu}Sv/h$, $1.75{\mu}Sv/h$, and $1.23{\mu}Sv/h$ in 100, 150, and 200 cm on abdomen. Transmittance was increased as the distance was expanded. As the distance was further, the radiation dose were reduced. When using shielder, the dose were reduced as one-forth of without shielder. The Radio technologists are exposed of radioactivity and there were limitations on reducing the distance with Therefore, the proper shielding will be able to decrease radiation dose to the technologists.