• Progress in Medical Physics
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• v.10 no.1
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• pp.1-7
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• 1999

In this study, the dose distributions of a $^{32}$ p uniform cylindrical volume source and a surface source, a pure $\beta$emitter, were calculated in order to obtain information relevant to the utilization of a balloon catheter and a radioactive stent. The dose distributions of $^{32}$ p were calculated by means of the EGS4 code system. The sources are considered to be distributed uniformly in the volume and on the surface in the form of a cylinder with a radius of 1.5 mm and length of 20 mm. The energy of $\beta$particles emitted is chosen at random in the $\beta$ energy spectrum evaluated by the solution of the Dirac equation for the Coulomb potential. Liquid water is used to simulate the particle transport in the human body. The dose rates in a target at a 0.5mm radial distance from the surface of cylindrical volume and surface source are 12.133 cGy/s per GBq (0.449 cGy/s per mCi, uncertainty: 1.51%) and 24.732 cGy/s per GBq (0.915 cGy/s per mCi, uncertainty: 1.01%), respectively. The dose rates in the two sources decrease with distance in both radial and axial direction. On the basis of the above results, the determined initial activities were 29.69 mCi and 1.2278 $\mu$Ci for the balloon catheter and the radioactive stent using $^{32}$ P isotope, respectively. The total absorbed dose for optimal therapeutic regimen is considered to be 20 Gy and the treatment time in the case of the balloon catheter is less than 3 min. Absorbed doses in targets placed in a radial direction for the two sources were also calculated when it expressed initial activity in a 1 mCi/ml volume activity density for the cylindrical volume source and a 0.1 mCi/cm$^2$ area activity density for the surface source. The absorbed dose distribution around the $^{32}$ P cylindrical source with different size can be easily calculated using our results when the volume activity density and area activity density for the source are known.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.253-266
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• 2016

In accordance with the classification system of radioactive waste in Korea, all the disused sealed radioactive sources (DSRSs) fall under the category of EW, VLLW or LILW, and should be managed in compliance with the restrictions for the disposal method. In this study, the management and disposal method are drawn in consideration of half-life of radionuclides contained in the source and A/D value (i.e. the activity A of the source dividing by the D value for the relevant radionuclide, which is used to provide an initial ranking of relative risk for sources) in addition to the domestic classification scheme and disposal method, based on the characteristic analysis and review results of the management practices in IAEA and foreign countries. For all the DSRSs that are being stored (as of March 2015) in the centralized temporary disposal facility for radioisotope wastes, applicability of the derivation result is confirmed through performing the characteristic analysis and case studies for assessing quantity and volume of DSRSs to be managed by each method. However, the methodology derived from this study is not applicable to the following sources; i) DSRSs without information on the radioactivity, ii) DSRSs that are not possible to calculate the specific activity and/or the source-specific A/D value. Accordingly, it is essential to identify the inherent characteristics for each of DSRSs prior to implementation of this management and disposal method.

• Journal of Radiation Protection and Research
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• v.23 no.4
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• pp.211-218
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• 1998

The distributional effects of radioactive materials on external gamma exposure have been analyzed. An approximate method for estimating external gamma dose given from an arbitrary distribution of radioactive material has been developed. The minimum gamma exposure given from a point source is shown at 0.07 MeV if the source to receptor distance is shorter than 10 m. But if the receptor to point source distance is longer than 20 m, gamma exposure rate increases monotonously according to the average gamma energy. For the analysis of the effects of volume source, we estimated the gamma dose given from different size of hemisphere in which radioactive materials are distributed uniformly. When the radius of hemisphere is longer than 40 m, external gamma dose rate increases monotonously. The gamma dose rate given from the radioactive materials deposited on the ground shows the minimum value at 0.07 MeV in any case. The analysis shows that external gamma exposure is strongly dependent on the distribution of radioactive materials in the environment and gamma energy.

• Journal of the Korean Society of Radiology
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• v.15 no.6
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• pp.821-831
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• 2021

In order to calibrate energy and efficiency using the PENELOPE Code, a PENELOPE simulation was performed using a volume source. Here, we want to verify peak efficiency and usefulness by performing simultaneous measurement and correction. calculate the coincident sum correction for all volumes, first subdivide the volumes of the cylinder and the four Marinelli beakers into three heights again. Therefore, the simultaneous measurement correction coefficient in three areas and the simultaneous measurement correction coefficient for the entire volume source are calculated as output. At low energies, the j value for each source volume (50-300 ml) is small and increases significantly in the high energy range. Simulation results showed good agreement within 2.5% for all source volumes except for 50 ml and 300 ml, which were up to 4%. This means that the correction for the simultaneous measurement effect during measurement is effective. In addition. Based on this, it can be confirmed that there is an advantage to improve the detection efficiency when measuring various sources and environmental samples.

• Composites Research
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• v.28 no.5
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• pp.311-315
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• 2015

Measuring fiber volume fraction properly is very important in designing composite materials because the fiber volume fraction mainly determines mechanical and thermal properties. Conventional Ignition methods are effective for ceramic fiber reinforcing composite materials. However, these methods are not proper for applying to carbon fiber reinforcing composites because of the venerable characteristic against oxidation of carbon fiber. In the research, fiber volume fraction of carbon fiber composites was obtained by a thermogravimetric analysis considering oxidation characteristic of the carbon fiber and the method was compared and verified with the results from microscopic cross section images.

• Journal of the Korean Society for Nondestructive Testing
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• v.17 no.3
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• pp.151-155
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• 1997

In this research, nondestructive test using a radioisotope, $^{241}Am$ gamma-ray, was accomplished in order to evaluate the fiber volume fraction of the accumulated composite layers such as glass fiber/epoxy and carbon fiber/epoxy. Attenuation coefficients of the fiber and resin were measured respectively by NaI(T1) detector The fibers volume fraction was measured for various thickness of composite layers between 2 and 20mm. Fiber volume fraction of the composite layers were also measured for various amount of fibers. The experimental errors from nondestructive test using gamma-ray were in the range of ${\pm}1{\sim}2.5%$ in comparison with those from observation by optical microscopy. By selecting the optimum energy and activity of radioisotope, this method can provide a new means for the evaluation of the fiber volume fraction.

• Composites Research
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• v.32 no.5
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• pp.270-278
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• 2019

In this study, the transverse mechanical properties of the unidirectional fiber reinforced composite modeled with fiber, matrix, and interphase is predicted with the representative volume elements and is calibrated by adjusting the properties and thickness of the interphase by referring to the test results. While the conventional representative volume elements modeled with fiber and matrix shows high predictive accuracy for the longitudinal mechanical properties, but it shows some deviations in the transverse mechanical properties. In order to compensate such gaps, the interphase region is employed, and its mechanical properties are adjusted to improve the prediction accuracy according to various elastic modulus, thickness, and strength parameters. As a result, the deviation of the transverse elastic modulus and strength is reduced significantly similar to the test results of the unidirectional composites with the accuracy of the longitudinal mechanical properties preserved.

• Journal of the Korean Society of Radiology
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• v.7 no.5
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• pp.333-338
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• 2013

Generally, it's known fact that intensity of radioactivity satisfies inverse-square law. However, the law was dissatisfied with practical experiment because of limited shape of scintillation detector. Especially, in the case of near distance between the surface of detector and the radioactive source, the difference grows larger. In the present study, reason of this difference was confirmed by experiment with $2^{{\prime}{\prime}}{\times}2^{{\prime}{\prime}}{\phi}$ NaI(Tl) scintillation detector and $^{60}Co$(1.174 MeV, 1.333 MeV)and $^{137}Cs$(0.662 MeV) gamma ray sources. From the experiment, the correction coefficient was obtained with gamma ray detection efficiency and geometrical volume. In the result of the present study, the efficiency difference of the detector was corrected with the coefficient. In the present result, we obtained that the inverse-square law experiment have to consider the efficiency and geometrical value of the detector.

• Proceedings of the KIEE Conference
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• 2002.07c
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• pp.1458-1459
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• 2002

PE(Polyethylene)와 PP(Polypropylene)의 sheet 에 $Co^{60}{\gamma}$-ray 선원을 실온, 대기 중에서 10 kGy, 30 kGy, 50 kGy의 방사선량으로 조사시키고 습도 분포에 따른 표면 및 체적 저항을 측정한 결과 방사선이 조사되지 않은 시료와 방사선이 조사된 시료와의 저항 값은 큰 차이를 보이지 않았으나 습도 효과에 따른 변화는 습도 분포가 높을수록 저항 값은 상대적으로 낮게 나타났다.

• 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.