• Journal of the Korean Society of Radiology
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• v.16 no.4
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• pp.471-476
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• 2022

The purpose of this study was to study the probability of cancer occurrence due to photoneutron dose exposure of the colon and thyroid gland, which are normal organs, in 3D CRT, IMRT 5 portals, and IMRT 9 portals, which are radiotherapy methods for prostate cancer. The total prescribed dose for prostate cancer was 6600 cGy, 220 cGy per dose, and 30 divided irradiations were applied for the total number of times. After setting up the Rando phantom on the treatment table (couch) of the medical linear accelerator used in the experiment, an optically stimulated luminescence albedo neutron dosimeter was placed on the corresponding area of the large intestine and thyroid gland of the phantom for measurement. During 3D CRT of prostate cancer, the probability of secondary cancer due to photoneutron dose to the colon and thyroid gland, which are normal organs, was 1.8 per 10,000 people. And IMRT 5 portals were 8.7 per 10,000 people, which was about 5 times larger than 3D CRT. IMRT 9 portals derived the result that there is a probability that 1.2 people per 1,000 people will develop cancer. Based on this study, the risk of secondary radiation exposure due to the dose of photoneutrons generated during radiation therapy is studied, and it is thought that it will be used as useful data for radiation protection in relation to the stochastic effect of radiation in the future.

• The Bulletin of The Korean Astronomical Society
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• v.37 no.2
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• pp.120.1-120.1
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• 2012

The Lunar Reconnaissance Orbiter (LRO) launched on June 16, 2009 has six experiments including of the Cosmic Ray Telescope for the Effects of Radiation (CRaTER) onboard. The CRaTER instrument characterizes the radiation environment to be experienced by humans during future lunar missions. The CRaTER instrument measures the effects of ionizing energy loss in matter specifically in silicon solid-state detectors due to penetrating solar energetic protons (SEP) and galactic cosmic rays (GCRs) after interactions with tissue-equivalent plastic (TEP), a synthetic analog of human tissue. The CRaTER instrument houses a compact and highly precise microdosimeter. It measures dose rates below one micro-Rad/sec in silicon in lunar radiation environment. Forbush decrease (FD) event is the sudden decrease of GCR flux. We use the data of cosmic ray and dose rates observed by the CRaTER instrument. We also use the CME list of STEREO SECCHI inner, outer coronagraph and the interplanetary CME data of the ACE/MAG instrument.We examine the origins and the characteristics of the FD-like events in lunar radiation environment. We also compare these events with the FD events on the Earth. We find that whenever the FD events are recorded at ground Neutron Monitor stations, the FD-like events also occur on the lunar environments. The flux variation amplitude of FD-like events on the Moon is approximately two times larger than that of FD events on the Earth. We compare time profiles of GCR flux with of the dose rate of FD-like events in the lunar environment. We figure out that the distinct FD-like events correspond to dose rate events in the CRaTER on lunar environment during the event period.

• Journal of radiological science and technology
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• v.45 no.1
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• pp.57-67
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• 2022

CZT detectors, which are compound semiconductors that have been widely used recently for gamma-ray detection purposes, are difficult to detect neutrons because direct interaction with them does not occur unlike gamma-rays. In this paper, a method of detecting and determining energy levels (fast neutrons and thermal neutrons) of neutrons, in addition of identifying energy and nuclide of gamma-rays, and evaluating gamma dose rates using a CZT semiconductor detector is described. Neutrons may be detected by a secondary photoelectric effect or compton scattering process with a characteristic gamma-ray of 558.6 keV generated by a capture reaction (¹¹³Cd + ¹n → ¹¹⁴Cd + 𝛾) with cadmium (Cd) in the CZT detector. However, in the case of fast neutrons, the probability of capture reaction with cadmium (Cd) is very low, so it must be moderated to thermal neutrons using a moderator and the material and thickness of moderator should be determined in consideration of the portability and detection efficiency of the equipment. Conversely, in the case of thermal neutrons, the detection efficiency decreases due to shielding effect of moderator itself, so additional CZT detector that do not contain moderator must be configured. The CZT detector that does not contain moderator can be used to evaluate energy, nuclide, and gamma dose-rate for gamma-rays. The technology proposed in this paper provides a method for detecting both neutrons and gamma-rays using a CZT detector.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.584-589
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• 2003

The hot cell facility which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ and $1.01{\times}10^{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ and $7.88{\times}10^{-2}$ mSv/h, respectively The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources, and penetration and toboggan will be partly reinforced by lead shield.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.43-49
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• 2022

The purpose of this study is to evaluate the radiation effect on damage when the external shield of the spent nuclear fuel transport cask is damaged due to impact as the cause of an unexpected accident. The neutron and gamma-ray intensities and spectra are calculated using the ORIGEN-Arp module in the SCALE 6.2.4 code package⁽¹⁾ and then using MCNP6.2⁽²⁾ code calculate the dose rate. In order to evaluate the radiation dose according to the size of damage caused by external impact, various sized holes of 0.3~13.7% are assumed in the outer shield of the cask to evaluate the sensitivity to the dose. In the case of radiation source leakage, damage to the nuclear fuel assembly is assumed to be up to 6% based on overseas test cases. When only the outer shield is damaged, the maximum surface dose is calculated as 3.12E+03 mSv/hr. However, if the radiation source is leaked due to damage to the nuclear fuel assembly, it becomes 7.00E+05 mSv/hr which is about 200 times greater than the former case.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.378-390
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• 2023

The SPES (Selective Production of Exotic Species) facility, currently under development at Legnaro National Laboratories of INFN, aims at the production of intense RIB (Radioactive Ion Beams) employing the Isotope Separation On-Line (ISOL) technique for interdisciplinary research. The radioactive isotopes of interest are produced by the interaction of a multi-foil uranium carbide target with a 40 MeV 200 μA proton beam generated by a cyclotron proton driver. The Target Ion Source (TIS) is the core of the SPES project, here the radioactive nuclei, mainly neutron-rich isotopes, are stopped, extracted, ionized, separated, accelerated and delivered to specific experimental areas. Due to efficiency reasons, the TIS unit needs to be replaced periodically during operation. In this highly radioactive environment, the employment of autonomous systems allows the manipulation, transport, and storage of the TIS unit without the need for human intervention. A dedicated remote handling infrastructure is therefore under development to fulfill the functional and safety requirement of the project. This contribution describes the layout of the SPES target area, where all the remote handling systems operate to grant the smooth operation of the facility avoiding personnel exposure to a high dose rate or contamination issues.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.1-11
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• 2004

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations peformed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}, 2.99{\times}10^{-3} and 7.88{\times}10^{-2}$ mSv/h, respectively, The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

• Proceedings of the Korea Crystallographic Association Conference
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• 2003.05a
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• pp.22-22
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• 2003

원자로에서 핵분열에 의해 생성된 고에너지 중성자는 감속재를 통해 열평형에 의해 에너지가 낮춰져 통계적 분포, 즉 Maxwell-Boltzman 운동에 따른 에너지 스펙트림을 갖게 된다. 중성자 산란장치는 통상 단색빔을 이용하므로 단색기(monochiomator)를 통해 이 분포에서 특정 파장의 중성자빔을 인출, 즉 단색화한다. 이때 단색기는 각각의 중성자 산란장치에 사용할 수 있는 특정 파장의 중성자빔을 인출하면서도, 파장의 퍼짐을 적절하게 조절하여 높은 중성자속(neutron flux)을 가지며 분해능도 또한 좋아야 한다. 전통적으로 많이 사용하는 단색화 방법은 결정의 내부결함을 유도하여 만든 모자익(mosaic) 결정을 이용하는 것이다. 이 방법은 특정 파장을 얻으면서도 좋은 분해능과 높은 중성자속을 갖는 모자익 결정을 만들기가 어렵고, 한번 결정된 단색기의 특성을 바꿀 수 없는 단점이 있다. 1980년대부터 몇몇 그룹이 거의 완전하게 성장된 단결정 슬랩을 미세하게 구부려서 탄성변형을 주어 effective 모자익 구조를 발생시킨 '구부린 완전결정(bent perfect crystal, BPC)' 단색기를 개발하여 특정 목적에 활용하는 시도를 하였다. BPC 단색기는 단색화된 중성자빔을 집속(focusing)할 수 있으며, 결정의 구부림 정도를 조절하고 배치 기하를 바꿈으로써 다양한 특성을 갖는 단색빔을 얻을 수 있는 장점이 있다. 이렇게 단색기의 기하학적 변수를 조절함으로써 회절빔의 집속도와 분해능을 조절할 수 있어서 잔류응력 측정이나 단결정 회절 및 집합조직 측정장치 등에 적용할 수 있다. 본 연구에서는 BPC 단색기의 원리와 여러 배치기하에 따른 빔의 특성을 소개하고자 한다.빔이 시료와 상호 작용하는 면적과 상호작용하지 않을 때의 빔을 회절모드에서 faraday cup으로 측정한 빔전류로 부터 계산하였다. Gibbsite에 대한 전자빔 조사 시 1분 이내에 급격한 Hydroxyl Ion(OH-)의 이탈로 인해 Cibbsite의 구조는 거시적 비정질화가 되며 시간증가에 따라 χ-alumina → ν-alumina → σ-alumina or δ-alumina의 순으로 상전이를 겪는다. 전자빔 조사 시 관찰된 회절자료의 가시적 변화를 통해 illumination angle 1.25mrad(Dose rate : 334 × 10³ e/sup -//sec·n㎡)일 경우 약 3초 이내에 비정질화가 시작됨을 알 수 있었고 이는 약 1 × 10/sup 6/ e/sup -//sec·n㎡ 의 전자선량에 해당되며 이를 기준으로 각각의 illumination angle에 대한 임계전자선량을 평가할 수 있었다. 실질적으로 Cibbsite와 같은 무기수화물의 직접가열실험 시 전자빔 조사에 의해 야기되는 상전이 영향을 배제하고 실험을 수행하려면 illumination angle 0.2mrad (Dose rate : 8000 e/sup -//sec·n㎡)이하로 관찰하고 기록되어야 함을 본 자료로부터 알 수 있었다.운동횟수에 의한 영향으로써 운동시간을 1일 6시간으로 설정하여, 운동횟수를 결정하기 위하여 오전, 오후에 각 3시간씩 운동시키는 방법과 오전부터 6시간동안 운동시키는 두 방법을 이용하여 품질을 비교하였다. 각 조건에 따라 운동시킨 참돔의 수분함량을 나타낸 것으로, 2회(오전 3시간, 오후 3시간)에 나누어서 운동시키기 위한 육의 수분함량은 73.37±2.02%를 나타냈으며, 1회(6시간 운

• Journal of Radiation Protection and Research
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• v.48 no.4
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• pp.184-196
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• 2023

Background: Radiation protection is crucial in various fields due to the harmful effects of radiation. Shielding is used to reduce radiation exposure, but gamma radiation poses challenges due to its high energy and penetration capabilities. Materials and Methods: This work investigates the radiation shielding properties of polyvinylidene fluoride (PVDF) samples containing different weight fraction of tungsten carbide (WC), tungsten trioxide (WO₃), and tungsten disulfide (WS₂). Parameters such as the mass attenuation coefficient (MAC), half-value layer (HVL), mean free path (MFP), effective atomic number (Z_eff), and macroscopic effective removal cross-section for fast neutrons (Σ_R) were calculated using the Phy-X/PSD software. EpiXS simulations were conducted for MAC validation. Results and Discussion: Increasing the weight fraction of the additives resulted in higher MAC values, indicating improved radiation shielding. PVDF-xWC showed the highest percentage increase in MAC values. MFP results indicated that PVDF-0.20WC has the lowest values, suggesting superior shielding properties compared to PVDF-0.20WO₃ and PVDF-0.20WS₂. PVDF-0.20WC also exhibited the highest Z_eff values, while PVDF-0.20WS₂ showed a slightly higher increase in Z_eff at energies of 0.662 and 1.333 MeV. PVDF-0.20WC has demonstrated the highest Σ_R value, indicating effective shielding against fast neutrons, while PVDF-0.20WS₂ had the lowest Σ_R value. The Monte Carlo N-Particle Transport version 5 (MCNP5) simulations showed that PVDF-xWC attenuates gamma radiation more than pure PVDF, significantly decreasing the dose equivalent rate. Conclusion: Overall, this research provides insights into the radiation shielding properties of PVDF mixtures, with PVDF-xWC showing the most promising results.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2317-2323
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• 2024

Medical linear accelerators with an energy of 8 MV or higher are radiated owing to photonuclear reactions and neutron capture reactions. It is necessary to quantitatively evaluate the concentration of radioactive isotopes when replacing or disposing them. HPGe detectors are commonly used to identify isotopes and measure radioactivity. However, because the detection efficiency is generally calibrated using a standard material with a density of 1.0 g/cm³, a self-absorption effect occurs if the density of the measured material is high. In this study, self-absorption correction factors were calculated for tungsten, lead, copper, and SUS-303, which are the main materials of medical linear accelerator head parts, for each gamma-ray energy using MCNP 6.2 code. The self-absorption effect was more pronounced as the energy of the emitted gamma rays decreased and the density of the measured materials increased. These correction factors were applied to the radioactivity measurements of the in-built and portable HPGe detectors. Furthermore, compared to the surface dose rate measured by the survey meter, the accuracy of the measurements of radioactivity improved by an average of 124.31 and 100.53 % for inbuilt and portable HPGe detectors, respectively. The results showed a good agreement, with an average difference of 3.70 and 5.24 %.