• Journal of radiological science and technology
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• v.45 no.6
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• pp.523-529
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• 2022

Gamma spectroscopy analysis is widely used for radioactivity analysis, and various factors are required for radioactivity calculations. Among the factors, K3 for each sample significantly influences the results. The previous methods of correcting the self-absorption effect include a computational simulation method and a method that requires making a CRM(certified reference material) identical to the sample medium. However, the above methods have limitations when used in small institutions because they require specialized program utilization skills or high manufacturing costs and large facilities. The aim of this study is to develop a method that can be easily and rapidly applied to radioactivity analysis. After filling the beaker with water, we placed the radiation source in a uniform position and used the measured value as the benchmark. Next, a correction factor was derived based on the difference in the radiation source count of the benchmark and the identically measured sample. For the radiation source, Eu-152, which emits a broad range of energy within the measurement range of gamma rays, and Cs-134 and Cs-137, which are indicator nuclides in environmental radiation analysis, were used. The sample was selected within the density range of 0.26-2.11 g/cm³, and the correction factor was derived by calculating the count difference of each sample compared to the reference value of water. This study presents a faster and more convenient method than the existing research methods for determining the self-absorption effect correction, which has become increasingly necessary.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.201-209
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• 2016

The object of this study was to evaluate the separation distance from a temporary storage facility satisfying the dose criteria. The calculation of ambient dose rates took into account cover soil thickness, facility size, and facility type by using MCNPX code. Shielding effects of cover soil were 68.9%, 96.9% and 99.7% at 10 cm, 30 cm and 50 cm respectively. The on-ground type of storage facility had the highest ambient dose rate, followed by the semi-ground type and the underground type. The ambient dose rate did not vary with facility size (except $5{\times}5{\times}2m\;size$) due to the self-shielding of decontamination waste in temporary storage. The separation distances without cover soil for a $50{\times}50{\times}2m\;size$ facility were evaluated as 14 m (minimum radioactivity concentration), 33 m (most probably radioactivity concentration), and 57 m (maximum radioactivity concentration) for on-ground storage type, 9 m, 24 m, and 45 m for semi-underground storage type, and 6 m, 16 m, and 31 m for underground storage type.

• 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.

• Journal of rehabilitation welfare engineering & assistive technology
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• v.11 no.4
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• pp.331-338
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• 2017

In this study, a portable x-ray generator was manufactured and a real-time image acquisition system was constructed using the image intensifier from the generated generator. We have developed a real - time position error verification system that can verify whether the artificial joint position is different from the initial image from the acquired image. The template image of the region of interest is extracted from the reference image using the pattern matching technique and compared with the image to be compared. As a result, It is shown that real - time position error verification is achieved by displaying the difference angle. This system is portable type, has a self-shielding facility, and the output of the irradiation device can be manufactured in a small size of 1kw and can be used as a portable type. In case of emergency patients in the non-destructive field for industrial use, It has proved effective for use in small areas such as feet.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.378-383
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• 2016

Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and $19.1g{\cdot}cm^{-3}$ density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.02a
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• pp.474-474
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• 2010

현재 반도체시장의 확장으로 인해서 기존의 300mm 웨이퍼에서 450mm의 웨이퍼를 사용하는 공정으로 변화하는 추세이다. 450mm 웨이퍼로 대면적 화되면서 기존 300mm 공정 때보다 훨씬 효율적인 플라즈마 소스 즉, 고밀도이고, 고균등화(high uniformity) 플라즈마 소스를 필요로 한다. 본 논문에서는 고밀도 플라즈마 소스인 유도 결합형 플라즈마(Inductively Coupled Plasma ; ICP)에 축 방향의 약한 자기장을 인가시킨 자화된 유도결합형 플라즈마(Magnetized Inductively Coupled Plasma : MICP)[1]를 제안하여 기존 ICP와의 차이점을 살펴보았다. 실험 방법으로 레이저 유기 형광법(Laser Induced Fluorescence : LIF)[2]을 이용하여 플라즈마 쉬스(Sheath) 내의 전기장을 외부 자기장의 변화에 따라 높이별로 측정하고 그 결과로부터 쉬스의 전기적 특성을 살펴보았다. 플라즈마의 특성상 탐침이나 전극에 전압을 인가하면 그 주위로 디바이 차폐(Debye Shielding)현상이 일어나서 플라즈마 왜곡이 일어난다. 그렇기에 플라즈마, 특히 플라즈마 쉬스의 특성을 파악하기 위해서 레이저라는 기술을 사용하였다. 레이저는 고가의 장비이고 그 사용에 많은 경험지식(know-how)를 필요로 하지만 플라즈마를 왜곡시키지 않고, 플라즈마의 밀도, 온도, 전기장 등 많은 상수(parameter)들을 얻어 낼 수 있다. 또한 3차원적으로 높은 분해능을 가지고 있는 장점이 있다. 강한 전기장이 있는 곳에서 입자들의 고에너지 준위가 전기장의 세기에 비례하여 분리되는 Stark effect[3] 이론을 이용하여 플라즈마 쉬스내의 전기장을 측정하였다. 실험은 헬륨가스 700mTorr 압력에서 이루어졌다. 기판의 파워를 50W에서 300W까지 변화시키면서 기판에 생기는 쉬스의 전기장의 변화를 살펴보았고, 자기장을 인가한 후 동일한 실험을 하여 자기장의 유무에 따른 플라즈마 쉬스의 전기장 변화를 살펴보았다. 실험결과 플라즈마 쉬스의 전기장의 변화는 기판의 파워와 플라즈마 밀도에 크게 의존함을 알았다. 기판의 파워가 커질수록 쉬스의 전기장은 커지고, 기판에 생기는 Self Bias Voltage역시 음의 방향으로 커짐을 확인 하였다. 또한 자기장을 걸어주었을 경우 쉬스의 두께가 얇아짐으로써 플라즈마의 밀도가 증가했음을 확인 할 수 있었다.

• The Korean Journal of Nuclear Medicine
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• v.39 no.1
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• pp.1-8
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• 2005

This paper is described on the development of KOTRON-13 and recent status of PET cyclotron by commercial cyclotron companies. KIRAMS has developed medical cyclotron which is KIRAMS-13. Samyoung Unitech produces KOTRON-13 with transfered technology by KIRAMS. As a part of Regional Cyclotron Installation Protect, KOTRON-13 cyclotrons and $[18F]FDG$ production modules are being installed at regional cyclotron centers in Korea. The medical concern with radiation technology has been growing for the last several years. Early cancer diagnosis through the cyclotron and PET-CT have been brought to public attention by commercial cyclotron models in the world. The new commercial cyclotron models are introduced compact low energy cyclotrons developed by CTI, GE, Sumitomo in recent. It produces different short-lived radioisotopes, such as $[^{18}F],\;[^{11}C],\;[^{13}N]\;and\;[^{15}O]$. For the better reliability acceleration particle is proton only. The characteristics of new model cyclotrons are changed to lower energy corresponding to less 13 MeV. New models have self-shielding and low power consumption. Design criteria for the different types of commercial cyclotrons are described with reference to hospital demands.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.549-555
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• 2017

A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2~3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

• Journal of the Korean Society of Radiology
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• v.13 no.1
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• pp.73-80
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• 2019

Neutrons are generated by the nuclear reaction, which is absorbed into the concrete wall and causes the activation during cyclotron operation. The purpose of this study is to investigate the effect of neutron activation and radiative concrete on concrete type. This experiment used Monte Carlo simulation and RESRAD model. The results of the experiment showed that the higher the content of Fe in concrete, the greater the shielding rate. The effect of $^{56}Fe(n,\;2np)^{54}Mn$ reaction on workers is also increased. However, radioactive nuclides have low activity and have very low impact on workers. Radioactive concrete should be treated as general wastes with less than its self-disposal tolerance level, and it should be recycled to the surface such as road repair rather than landfill to minimize the effect of $^{14}C$.

• Journal of the Institute of Convergence Signal Processing
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• v.21 no.4
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• pp.195-201
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• 2020

This study studied scoliosis, one of the most common modern diseases caused by lifestyle patterns of office workers sitting in front of computers all day and modern people who use smart phones frequently. Scoliosis is a typical complication that takes more than 80% of the nation's total population at least once. X-ray are used to test for these complications. X-ray, a non-destructive testing method that allows scoliosis to be easily performed and filmed in various areas such as the chest, abdomen and bone without contrast agents or other instruments. We uses NI DAQ to miniaturize digital X-ray imaging devices and image intensifier in self-shielding housing with Vision Assistant for drawing lines to the top and the bottom of the spine to acquire angles, i.e. curvature in real-time. In this way, the research was conducted to see scoliosis patients and their condition easily and to help rapid treatment for solving the problem of posture correction in modern people.