• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.282-285
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• 2016

Background: To optimize the photon transmission efficiency of light guides used in scintillation detectors, LightTools code, which can construct and track light, was used to analyze photon transmission effectiveness with respect to light guides thickness. Materials and Methods: This analysis was carried out using the commercial light guide, N-BK 7 Optical Glass by SCHOTT, as a model for this study. The luminous exitance characteristic of the LYSO scintillator was used to analyze the photon transmission effectiveness according to the thickness of the light guide. Results and Discussion: The results of the simulations showed the effectiveness of the photon transmission according to the thickness of the light guide, which was found to be distributed from 13.38% to 33.57%. In addition, the photon transmission efficiency was found to be the highest for light guides of 4 mm of thickness and a receiving angle of $49^{\circ}$. Conclusion: Through such simulations, it is confirmed that photon transmission efficiency depends on light guide thickness and subsequent changes in the internal angle of reflection. The aim is to produce an actual light guide based on these results and to evaluate its performance.

• 대한방사선기술학회지:방사선기술과학
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• 제44권6호
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• pp.629-633
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• 2021

This article is designed to look into the radiation exposure dose to each body part and the shielding effect for workers using an additional shielding to reduce their radiation exposured by scattering radiation which is generated in a space between the operating table and lead curtain during interventional radiology(IR) procedures. After placing a human phantom on the table of SIEMENS' angiography machine, the following measurements were taken, depending on the presence of an additional shield of lead equivalent of 0.25 mmPb, manufactured for this purpose: dose to gonad, dose to an area where the personal dosimeter is placed, and dose to an area of eye lens is located. An ion chamber(chamber volume 1,800 cc) was utilized to measure scattering radiation. The two imaging tests were carried out as follows: fluoroscopy of the abdomen (66 kV, 100 mA, 60 seconds) and of the head (70 kV, 65 mA, 60 seconds); and digital subtraction angiography(DSA) of the abdomen (67 kV, 264 mA, 20 seconds) and of the head (79 kV, 300 mA, 20 seconds). In all the experiments, the shielding efficiency of the gonad position was the largest at 59.8%. In case an additional shielding was used as protection against scattering radiation that came through the operating table and the lead curtain during an IR, the radiation shielding efficiency was estimated to be up to 59.8%, leading to a conclusion that its presence may effectively reduce the radiation exposure dose of medical staffs.

• 한국환경과학회지
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• 제11권10호
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• pp.1117-1123
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• 2002

Coating ceramic balls activate water molecular in water as radiate far-infrared radiation of high efficiency, and then prevent formation of scale and corrosion of pipe. Therefore, but only keep away drop of heat exchange efficiency of boiler, not also remove formed scale. As a result, pipe of boiler has clean and has form thin protection film in inner of pipe. Also, the water treated by rust preventing system using the ceramic balls, that is harmless, tasteless, odorless, and non-toxicity in the human body, and it can use drinking water. This rust preventing system can save energy and protect environment.

• Journal of Radiation Protection and Research
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• 제45권1호
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• pp.35-44
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• 2020

Background: Cadmium zinc telluride (CZT) is a promising material because of a high detection efficiency, good energy resolution, and operability at room temperature. However, the cost of CZT dramatically increases as its size increases. In this study, to achieve a large effective volume with relatively low cost, an array structure comprised of individual virtual Frisch-grid CZT detectors was proposed. Materials and Methods: The prototype consisted of 2 × 2 CZTs, a holder, anode and cathode printed circuit boards (PCBs), and an application-specific integrated circuit (ASIC). CZTs were used and the non-contacting shielding electrode method was applied for virtual Frisch-grid effect. An ASIC was used, and the holder and the PCBs were fabricated. In the current system, because the CZTs formed a common cathode, a total of 5 channels were assigned for data processing. Results and Discussion: An experiment using ¹³⁷Cs at room temperature was conducted for 10 minutes. Energy and timing information was acquired and the depth of interaction was calculated by the timing difference between the signals of both electrodes. Based on obtained three-dimensional position information, the energy correction was carried out, and as a result the energy spectra showed the improvements. In addition, a Compton image was reconstructed using the iterative method. Conclusion: The virtual Frisch-grid CZT detector based on the array structure was developed and the energy spectra and the Compton image were successfully acquired.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.299-303
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• 2001

Using ultrasound techniques, the Korea Atomic Energy Research Institute has measured chest wall thicknesses of a group of male workers at the Korea Atomic Energy Research Institute. A site-specific biometric equation has been developed for these workers. Chest wall thickness is an important modifier on lung counting efficiency. These data have been put into the perspective of the ICRP recommended dose limits for occupationally exposed workers: 100 mSv in a 5-year period with a maximum of 50 mSv in anyone year. For measured chest wall thicknesses of 1.9 cm to 4.1 cm and a 30 min counting time, the achievable MDAs for natural uranium in the KAERI lung counter vary from 5.75 mg to 11.28 mg. These values are close to, or even exceed, the predicted amounts of natural uranium that will remain in the lung (absorption type M and S) after an intake equal to the Annual Limit on Intake corresponding to a committed dose of 20 mSv. This paper shows that the KAERI lung counter probably cannot detect an intake of Type S natural uranium in a worker with a chest wall thickness equal to the average value (2.7 cm) under routine counting conditions.

• Journal of Radiation Protection and Research
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• 제31권3호
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• pp.149-153
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• 2006

An ionization chamber is still widely used in many fields by virtue of its' simple operational characteristics and the possibility of its' various shapes. A parallel type of an ionization chamber for a steel sheet thickness measurement was designed and fabricated. High pure xenon gas, which was pressurized up to 6 atm, was chosen as a filling gas to increase the current response and sensitivity for a radiation. A high pressure gas system was also constructed. The active volume and the incident window size of the fabricated ionization chamber were $30\;cm^3\;and\;12\;cm^2$, respectively. Preliminary tests with a 25 mCi $^{241}Am$ gamma-ray source and evaluation tests in a standard X-ray field were performed. The optimal operation voltage was set from the results of the collection efficiency calculation by using an experimental two-voltage method. Linearity for a variation of the steel sheet thickness, which is the most important factor for an application during a steel sheet thickness measurement, was 0.989 in this study.

• Journal of Radiation Protection and Research
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• 제38권4호
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• pp.194-201
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• 2013

To investigate monitoring unit (MU) efficiency and plan quality of volumetric modulated arc therapy (VMAT) using flattening-filter free (FFF) photon beam in association with target size and location. A virtual patient was generated in Eclipse$^{TM}$ (ver. A10, Varian Medical Systems, Palo Alto, USA) treatment planning system. The length of major and minor axis in axial view was 50 cm and 30 cm, respectively. Cylindrical-shaped targets were generated inside that patient at the center (symmetric target) and in the periphery (asymmetric target, 7.5 cm away from the center of the patient to the right direction) of the virtual patient. The longitudinal length was 10 cm and the diameters were 2, 5, 10 and 15 cm. Total 8 targets were generated. RapidArc$^{TM}$ plans using TrueBeam STx$^{TM}$ were generated for each target. Two full arcs were used and the axis of rotation of the gantry was set to be at the center of the virtual patient. Total MU, homogeneity index (HI), target mean dose, the value of gradient measure and body mean dose were calculated. In the case of symmetric targets, averaged total MU of FFF plan was 23% and 19% higher than that of flattening filter (FF) plan when using 6 MV and 10 MV photons, respectively. The difference of HI, target mean dose, gradient measure and body mean dose between FF and FFF was less than 0.04, 2.6%, 0.1 cm and 2.2%, respectively. For the asymmetric targets, total MU of FFF plan was 21% and 32% was higher than that of FF when using 6 MV and 10 MV photons, respectively. The homogeneity of the target was always worse when using FFF than using FF. The maximum difference of HI was 0.22. The target mean dose of FFF was 3.2% and 4.1% higher than that of FF for the 6 MV and 10 MV, respectively. The difference of gradient measure was less than 0.1 cm. The body mean dose was higher when using FFF than FF about 4.2% and 2.8% for the 6 MV and 10 MV, respectively. No significant differences between VMAT plans of FFF beam and FF beam were observed in terms of quality of treatment plan. The HI was higher when using FFF 10 MV photons for the asymmetric targets. The MU was increased noticeably when using FFF photon beams.

• Journal of Radiation Protection and Research
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• 제47권4호
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• pp.181-194
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• 2022

Internal dosimetry is a discipline which brings together a set of knowledge, tools and procedures for calculating the dose received after incorporation of radionuclides into the body. Several steps are necessary to calculate the committed effective dose (CED) for workers or members of the public. Each step uses the best available knowledge in the field of radionuclide biokinetics, energy deposition in organs and tissues, the efficiency of radiation to cause a stochastic effect, or in the contributions of individual organs and tissues to overall detriment from radiation. In all these fields, knowledge is abundant and supported by many works initiated several decades ago. That makes the CED a very robust quantity, representing exposure for reference persons in reference situation of exposure and to be used for optimization and assessment of compliance with dose limits. However, the CED suffers from certain limitations, accepted by the International Commission on Radiological Protection (ICRP) for reasons of simplification. Some of its limitations deserve to be overcome and the ICRP is continuously working on this. Beyond the efforts to make the CED an even more reliable and precise tool, there is an increasing demand for personalized dosimetry, particularly in the medical field. To respond to this demand, currently available tools in dosimetry can be adjusted. However, this would require coupling these efforts with a better assessment of the individual risk, which would then have to consider the physiology of the persons concerned but also their lifestyle and medical history. Dosimetry and risk assessment are closely linked and can only be developed in parallel. This paper presents the state of the art of internal dosimetry knowledge and the limitations to be overcome both to make the CED more precise and to develop other dosimetric quantities, which would make it possible to better approximate the individual dose.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1421-1430
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• 2022

The computation of the solid angle and the detector efficiency is considering to be one of the most important factors during the measuring process for the radioactivity, especially the cylindrical γ-ray NaI(Tl) detectors nowadays have applications in several fields such as industry, hazardous for health, the gamma-ray radiation detectors grow to be the main essential instruments in radiation protection sector. In the present work, a generic numerical simulation method (NSM) for calculating the efficiency of the γ-ray spectrometry setup is established. The formulas are suitable for any type of source-to-detector shape and can be valuable to determine the full-energy peak and the total efficiencies and P/T ratio of cylindrical γ-ray NaI(Tl) detector setup concerning the truncated conical radioactive source. This methodology is based on estimate the path length of γ-ray radiation inside the detector active medium, inside the source itself, and the self-attenuation correction factors, which typically use to correct the sample attenuation of the original geometry source. The calculations can be completed in general by using extra reasonable and complicate analytical and numerical techniques than the standard models; especially the effective solid angle, and the detector efficiency have to be calculated in case of the truncated conical radioactive source studied condition. Moreover, the (NSM) can be used for the straight calculations of the γ-ray detector efficiency after the computation of improvement that need in the case of γ-γ coincidence summing (CS). The (NSM) confirmation of the development created by the efficiency transfer method has been achieved by comparing the results of the measuring truncated conical radioactive source with certified nuclide activities with the γ-ray NaI(Tl) detector, and a good agreement was obtained after corrections of (CS). The methodology can be unlimited to find the theoretical efficiencies and modifications equivalent to any geometry by essential sufficiently the physical selective considered situation.

• Journal of Radiation Protection and Research
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• 제12권2호
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• pp.19-27
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• 1987

황산망간 용액조장치의 $^{56}Mn\;{\gamma}$선 검출효율을 결정하는데 $^{56}Mn$용액의 방사능을 절대측정하는 것은 필수적이다 $^{56}Mn$시료를 제작하기 위하여 99.99%의 순도를 갖는 Mn금속조각 13.718mg되는 시료를 한국에너지연구소 TRIGA MARK-II 원자로의 중성자선속이 약 $10^{13}n/cm^2{\cdot}s$되는 열중성자장에서 12분간 조사시켰다. 중성자 방사화된 $^{56}Mn$금속시료를 0.1N-HCI 용액 50ml 용해시켜서 $^{56}Mn$시료를 제작하여 $4{\pi}{\beta}-{\gamma}$ 동시계수기술로 방사능을 측정한 결과 불확도 0.366%를 갖는 값으로서 1987년 10월 15일 0 시를 기준하여 408.070kBq/mg을 얻었다.