• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.419-424
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• 2004

Various radiation counters have been using to determine radioactivity of radwastes for disposal. A radiation counting system was set up using a radiation detector chosen in this study and its stability was investigated through the periodic determination of background and counting efficiencies in accordance with a quality control program to increase the confidence level. The average background level for the $\gamma$-spectrometer was 1.59 cps and the average counting level for the standard sample was 45248 Ops within $2{\sigma}$ confidence levels. The average alpha background level for the low background ${\alpha}{\beta}$ counting system was 0.31 cpm and the efficiency for alpha counting was 34.38%. The average beta background level for the ${\alpha}{\beta}$ counting system was 1,30 cpm and the efficiency for beta counting was 46.5%, The background level in the region of 3H and 14C for the liquid scintillation counting system was 2.52 and 3.31 cpm and the efficiency for alpha counting was 58.5 and 95.6%, respectively. The minimum detectable activity for the$\gamma$-spectrometer was found to be 3.2 Bq/$m\ell$ and 3.8 Bq/$m\ell$ for the liquid scintillation counter, and 20.5 and 23.0 Bq/$m\ell$, respectively for the $\alpha$ and $\beta$ counting system.

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.37-42
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• 2009

Whole body counters (WBCs) are used to monitor radiation workers for internal contamination of radionuclides at domestic nuclear power plants (NPPs). A WBC is a scintillation detector using sodium iodide (NaI) and provides the identification of inhaled radionuclide and the measurement of its internal radioactivity in a short time. However, it is often possible to estimate external contamination as internal contamination due to radionuclides attached to the skin of radiation workers and this leads to an excessively conservative estimation of radioactive contamination. In this study, several experiments using a WBC and the Korean humanoid phantom were performed to suggest the more systematic method of discrimination between external and internal contamination. Furthermore, a WBC geometry experiment was conducted to suggest the optimal WBC geometry in consideration of deposited areas inside the body for dominant radionuclides at NPPs. The procedure of measurement and estimation of internal radioactivity for radiation workers at NPPs was improved on the basis of experimental results. Thus, it is expected to prevent from estimating internal exposure dose conservatively owing to the application of accurate whole body counting program to NPPs.

• Progress in Medical Physics
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• v.27 no.1
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• pp.8-13
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• 2016

Dual energy cone-beam CT can distinguish two materials with different atomic compositions. The principle of dual energy cone-beam CT based on modulation layer is that higher energy spectrum can be acquired at blocked x-ray window. To evaluate the possibility of modulation layer based dual energy cone-beam CT, we analyzed x-ray spectrum for various thicknesses of modulation layers by Monte Carlo simulation. To compare with the results of simulation, the experiment was performed on prototype cone-beam CT for 50~100 kVp with CdTe XR-100T detector. As the result of comparing, the mean energy of energy spectrum for 80 kVp are well matched with that of simulation. The mean energy of energy spectrum for 80 and 120 kVp were increased as 1.67 and 1.52 times by 2.0 mm modulation layer, respectively. We realized that the virtual dual energy x-ray source can be generated by modulation layer.

• The Korean Journal of Nuclear Medicine Technology
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• v.18 no.2
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• pp.57-61
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• 2014

Purpose Iodine (I-131) is one of the most widely used radioactive isotopes for therapeutic in the field of nuclear medicine. Therapeutic I-131 capsule is made out of lead to shield high energy radiation. Accurate dosimetry is necessarily required to perform safe and effective work for relative workers. The Monte Carlo method is known as a method to predict the absorbed dose distribution most accurately in radiation therapy and many researchers constantly attempt to apply this method to the dose calculation of radiotherapy recently. This paper aims to calculate distance dependent and activity dependent therapeutic I-131 capsule using GEANT4. Materials and Methods Therapeutic capsules was implemented on the basis of the design drawings. The simulated dose was determined by generating of gamma rays of energy to more than 364 keV. The simulated dose from the capsule at the distance of 10 cm and 100 cm was measured and calculated in the model of water phantom. The simulated dose were separately calculated for each position of each detector. Results According to the domestic regulation on radiation safety, the dose at 10 cm and 100 cm away from the surface of therapeutic I-131 capsule should not exceed 2.0 mSv/h and 0.02 mSv/h, respectively. The simulated doses turned out to be less than the limit, satisfying the domestic regulation. Conclusion These simulation results may serve as useful data in the prediction of hands dose absorbed by I-131 capsule handling. GEANT4 is considered that it will be effectively used in order to check the radiation dose.

• Journal of radiological science and technology
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• v.41 no.3
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• pp.201-207
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• 2018

The purpose of this study is to evaluate the performance of a "stealth chamber" as a novel reference chamber for measuring percentage depth dose (PDD) and profile of 6, 8 and 10 MV photon energies. The PDD curves and dose profiles with fields ranging from $3{\times}3$ to $25{\times}25cm^2$ were acquired from measurements by using the stealth chamber and CC 13 chamber as reference chamber. All measurements were performed with Varian VitalBeam linear accelerator. In order to assess the performance of stealth chamber, PDD curves and profiles measured with stealth chamber were compared with measurement data using CC13 chamber. For PPDs measured with both chambers, the dosimetric parameters such as $d_{max}$ (depth of maximum dose), $D_{50}$ (PDD at 50 mm depth), and $D_{100}$ (PDD at 100 mm depth) were analyzed. Moreover, root mean square error (RMSE) values for profiles at $d_{max}$ and 100 mm depth were evaluated. The measured PDDs and profiles between the stealth chamber and CC13 chamber as reference detector had almost comparable. For PDDs, the evaluated dosimetric parameters were observed small difference (<1%) for all energies and field sizes, except for $d_{max}$ less than 2 mm. In addition, the difference of RMSEs for profiles at $d_{max}$ and 100 mm depth was similar for both chambers. This study confirmed that the use of stealth chamber for measuring commission beam data is a feasible as reference chamber for fields ranging from $3{\times}3$ to $20{\times}20cm^2$. Furthermore, it has an advantage with respect to measurement of the small fields (less than $3{\times}3cm^2$ field) although not performed in this study.

• Journal of Radiation Protection and Research
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• v.24 no.1
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• pp.39-43
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• 1999

Thermoluminescent (TL) response characteristics of a thin LiF:Mg,Cu,Na,Si Teflon detectors have been studied for use in beta radiation detection. The detectors were fabricated from a mixture of LiF:Mg,Cu,Na,Si phosphor and Teflon powder which was molded into a thin disk form of $50mg/cm^2$ thickness. These detectors were irradiated to beta fields of $^{147}Pm,\;^{204}Tl\;and\;^{90}Sr/^{90}Y$ sources with a covering of Kapton foil ($2mg/cm^2$) and photon irradiation was carried out with a $^{137}Cs$ source at the Korea Atomic Energy Research Institute (KAERI). Batch uniformity was estimated to be 4.7% and the beta dose response presented linear relationship from 0.1 mGy to 100 Gy. The beta energy responses of thin detectors normalized to $^{137}Cs$ were presented as 0.46, 1.09 and 1.06 for $^{147}Pm,\;^{204}Tl\;and\;^{90}Sr/^{90}Y$ beta rays, respectively. The evaluated values for angular responses were $0.93{\pm}0.03\;(^{147}Pm),\;0.94{\pm}0.04\;(^{204}Tl),\;and\;0.92{\pm}0.05\;(^{90}Sr/^{90}Y)$. The results satisfied well a proposed ISO Standard for beta ray dosimeters.

• Radiation Oncology Journal
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• v.17 no.3
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• pp.256-260
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• 1999

Purpose : An investigation has been carried out on the factors which affect the response reading of thermoluminescent dosimeters (TLD-100) loaded with thin material in high energy Photon. The aim of the study was to assess the energy response of TLD-100 to the therapeutic ranges of photon beam. Materials and Methods : In this technique, TLD-100 (abbreviated as TLD) chips and three different thin material (Tin, Gold, and Tissue equivalent plastic plate) which mounted on the TLD chip were used in the clinical photon beam. The thickness of each metal plates was 0.1 mm and TE plastic plate was 1 mm thick. These compared with the photon energy dependence of the sensitivities of TLD (normal chip), TLD loaded with Tin or Gold plate, for the photon energy range 6 MV to 15 MV, which was of interest in radiotherapy. Results : The enhancement of surface dose in the TLD with metal plate was clearly detected. The TLD chips with a Gold plate was found to larger response by a factor of 1.83 in 10 MV photon beam with respect to normal chip. The sensitivity of TLD loaded with Tin was less than that for normal TLD and TLD loaded with Gold. The relative sensitivity of TLD loaded with metal has little energy dependence. Conclusion : The good stability and linearity with respect to monitor units of TLD loaded with metal were demonstrated by relative measurements in high energy Photon ($6\~15$ MV) beams. The TLD laminated with metals embedded system in solid water phantom is a suitable detector for relative dose measurements in a small beam size and surface dose.

• Journal of Radiation Protection and Research
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• v.17 no.2
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• pp.61-70
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• 1992

Photon, neutron and beta radiation fields were measured at PWR plants which are the representative types of nuclear power plant operated in Korea. The photon energy spectra were measured at locations in the auxiliary building during operation period and in the containment vessel(C/V) during shutdown period using a portable gamma spectrometer with a HPGe detector. The distribution of average energy was found to range from 440 to 780 keV in the C/V and from 280 keV to 760 keV in the auxiliary building, respectively. The average neutron energy measured at the five locations around the operation deck in the C/V in operation using a BMSS (Bonner Multi-Sphere Spectrometer) ranged from 20 keV to 210 keV. A computer code, BUNKI was used to unfold the spectrum. The beta energy spectra in the C/V and in the auxiliary building in annual outage were determined using 14 smear samples taken from the highly contaminated areas. The analysis showed that the representative corrosion product, $^{60}Co$ made main contribution to the beta energy field.

• The Korean Journal of Food & Health Convergence
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• v.4 no.1
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• pp.15-22
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• 2018

A radiation dosimeter is important to assess quality assurance (QA) of radiation therapy devices and to estimate the radiation dose in vivo dosimetry. Recently, optically stimulated luminescence detector (OSLD) is widely used in clinical filed. Therefore, the purpose of this study is to evaluate dose, energy, and angular dependence of OSLD and EBT3 film. The absorbed dose in clinical linear accelerator (Linac) beam is calibrated for dose per monitor unit (MU). Dose, energy, and angular dependence of OSLD and EBT3 film are estimated after the calibration procedure. The absorbed dose is measured at 50, 100, 150, and 200 cGy in an 6 MV X-ray beam for dose dependence. A dose of 150 cGy is delivered to OSLD and EBT3 film with 6 and 10 MV photon energies for energy dependence. For measurements of angular dependence, angular positions of gantry are $0^{\circ}{\pm}80^{\circ}$ with 6 MV at 150 cGy. The results of dose dependence is linear for OSLD and EBT3 film. For the results of energy dependence, errors were 0.39% and 0.03% for OSLD and EBT3 film, respectively. The results of dose for angular is decreased from $0^{\circ}$ to ${\pm}80^{\circ}$ for both OSLD and EBT3 film. When angle of $0^{\circ}$ is normalized to 1, and the dose is decreased to 60 and 66% at $80^{\circ}$ for OSLD and EBT3 film, respectively. Dose and energy dependence of OSLD and EBT3 film are measured within the recommendation of manufacturer. Angular dependence is increased from $0^{\circ}$ to ${\pm}80^{\circ}$ for OSLD and EBT3 film. The characteristics of OSLD and EBT3 film are similar and expected to useful for clinical field.

• Journal of Radiation Industry
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• v.10 no.4
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• pp.173-179
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• 2016

Coals and coal ashes, raw materials and by-products, in coal-fired power plants contain naturally occurring radioactive materials (NORM). They may give rise to internal exposure to workers due to inhalation of airborne particulates containing radioactive materials. It is necessary to characterize radioactivity concentrations of the materials for assessment of radiation dose to the workers. The objective of the present study was to analyze radioactivity concentrations of coals and by-products at four coal-fired plants in Korea. High purity germanium detector was employed for analysis of uranium series, thorium series, and potassium 40 in the materials. Radioactivity concentrations of $^{226}Ra$, $^{228}Ra$, and $^{40}K$ were $2{\sim}53Bq\;kg^{-1}$, $3{\sim}64Bq\;kg^{-1}$, and $14{\sim}431Bq\;kg^{-1}$ respectively in coal samples. For coal ashes, the radioactivity concentrations were $77{\sim}133Bq\;kg^{-1}$, $77{\sim}105Bq\;kg^{-1}$, and $252{\sim}372Bq\;kg^{-1}$ in fly ash samples and $54{\sim}91Bq\;kg^{-1}$, $46{\sim}83Bq\;kg^{-1}$, and $205{\sim}462Bq\;kg^{-1}$ in bottom ash samples. For flue gas desulfurization (FGD) gypsum, the radioactivity concentrations were $3{\sim}5Bq\;kg^{-1}$, $2{\sim}3Bq\;kg^{-1}$, and $22{\sim}47Bq\;kg^{-1}$. Radioactivity was enhanced in coal ash compared with coal due to combustion of organic matters in the coal. Radioactivity enhancement factors for $^{226}Ra$, $^{228}Ra$, and $^{40}K$ were 2.1~11.3, 2.0~13.1, and 1.4~7.4 for fly ash and 2.0~9.2, 2.0~10.0, 1.9~7.7 for bottom ash. The database established in this study can be used as basic data for internal dose assessment of workers at coal-fired power plants. In addition, the findings can be used as a basic data for development of safety standard and guide of Natural Radiation Safety Management Act.