• Journal of Radiation Protection and Research
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• v.7 no.1
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• pp.34-48
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• 1982

The prime purpose of this study is to realize an index quantity, absorbed dose index, defined by the ICRU for the characterization of ambient radiation level at any location for the purpose of radiation protection. The experiment has been designed to be carried out in two phases, namely, preliminary and main experiment. In the primary study a 30cm diameter sphere of polyethylene was used, while in the main experiment that of tissue equivalent material was fabricated and used. Both experiments were performed in the gamma-ray fields of $^{137}Cs\;and\;^{60}Co$, and in a neutron beam of thermal column of the TRIGA MARK-II research reactor. In the measurement of gamma-ray absorbed dose TLD-700 $(^{7}LiF)$ chips were used, and for the neutron dose both Au activation foils and TLD chips (TLD-600 $(^{6}LiF)$ and TLD-700 for the discrimination of gamma-ray contribution) were used. Theoretical assessment of the absorbed dose in the sphere phantom has been carried out in accordance with the Ehrlich's idea that deduced on the basis of Burlin's cavity theory in the case of gamma-ray irradiation. For the analysis of neutron dose fluence-KERMA rate conversion method was used. The explanation on the dose assessment is given in detail. Results obtained were numerically and statistically analyzed and the depth dose distributions are presented in the graphic forms with normalized values. In the concluding remarks, the possibility and difficulty of realizing the index quantity, including questions and problems to be solved are mentioned.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.273-280
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• 2014

The spectrum weighted responses of various detectors were calculated to provide guidance on the proper selection and use of survey instruments on the basis of their energy response characteristics on the neutron fields. To yield the spectrum weighted response, the detector response functions of 17 neutron-measuring devices were numerically folded with each of the produced calibration neutron spectra through the in-house developed software 'K-SWR'. The detectors' response functions were taken from the IAEA Technical Reports Series No. 403 (TRS-403). The reference neutron fields of 21 kinds with 2 spectra groups with different proportions of thermal and fast neutrons have been produced using neutrons from the $^{241}Am$-Be sources held in a graphite pile, a bare $^{241}Am$-Be source, and a DT neutron generator. Fluence-average energy ($E_{ave}$) varied from 3.8 MeV to 16.9 MeV, and the ambient-dose-equivalent rate [$H^*(10)/h$] varied from 0.99 to 16.5 mSv/h.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.275-283
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• 2019

Rare isotope beam facilities require shielding data in early stage of their design. There is much less shielding data on neutrons from the reactions between heavy ion beams and matter than the data on neutrons produced by protons. The purpose of the present work is to produce and thus increase the amount of shielding data on neutrons generated by high-energy heavy ion beams based on the RAON in-flight fragment facility. Calculations were performed with the computational Monte Carlo codes PHITS and MCNPX. The secondary neutron source terms were evaluated at 550 MeV/u for Ca, Kr, and Sn and at 400 MeV/u for U ions on a graphite target. Source terms and attenuation lengths were obtained by fitting the ambient dose equivalent inside an ordinary concrete shield.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.770-780
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• 2022

In 1993 the University of Extremadura initiated the design, construction and management of the Radiological Alert Network of Extremadura (RAREx). The goal was to acquire reliable near-real-time information on the environmental radiological status in the surroundings of the Almaraz Nuclear Power Plant by measuring, mainly, the ambient dose equivalent. However, the phased development of this network has been carried out from two points of view. Firstly, there has been an increase in the number of stations comprising the network. Secondly, there has been an increase in the number of monitored parameters. As a consequence of the growth of RAREx network, large data volumes are daily generated. To face this big data paradigm, software applications have been developed and implemented in order to maintain the indispensable real-time and efficient performance of the alert network. In this paper, the description of the current status of RAREx network after 30 years of design and performance is showed. Also, the performance of the graphing software for daily assessment of the registered parameters and the automatic on real time warning notification system, which aid with the decision making process and analysis of values of possible radiological and non-radiological alterations, is briefly described in this paper.

• International Journal of Safety
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• v.11 no.1
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• pp.33-36
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• 2012

Examples and experiences of risk management on radiation under prolonged exposure situation are shown. The accident of the Fukushima dai-ichi nuclear power plant after the great east Japan earthquake (11 March, 2011) elevates background level of environmental radiation around the east Japan. For example, ambient dose equivalent rate around Tohkatsu area next to Tokyo located about 200 km-south from the plant, is about 0.1-0.6 micro-Sv $h^{-1}$ mainly due to $^{134}Cs$ and $^{137}Cs$ falling on the ground soil. This level is about double or up to ten times higher than the genuine natural level around the area. International Commission on Radiological Protection (ICRP) recommends how to face the existing exposure situation; that is the prolonged exposure situation. Referring to ICRP's reports and/or related international/domestic documents, we have been discussing and acting to gain public's safety and relief, who have a possibility to be exposed to prolonged lower-dose radiation. Here, we introduce our several experiences on risk management, especially focusing on risk communication, radiation education to public, and stakeholder involvements into making decision in local governments on radiation protection, relating to the accident.

• International Journal of Safety
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• v.10 no.2
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• pp.6-9
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• 2011

Examples and experiences of risk management on radiation under prolonged exposure situation are shown. The accident of the Fukushima dai-ichi nuclear power plant after the great east Japan earthquake (11 March, 2011) elevates background level of environmental radiation around the east Japan. For example, ambient dose equivalent rate around Tohkatsu area next to Tokyo located about 200 km-south from the plant, is about 0.1-0.6 micro-Sv $h^{-1}$ mainly due to $^{134}Cs$ and $^{137}Cs$ falling on the ground soil. This level is about double or up to ten times higher than the genuine natural level around the area. International Commission on Radiological Protection (ICRP) recommends how to face the existing exposure situation; that is the prolonged exposure situation. Referring to ICRP's reports and/or related international/domestic documents, we have been discussing how to manage this situation and acting to gain safety and relief of public, who have a possibility to be exposed to prolonged lower-dose radiation. Here, we introduce our several experiences on risk management, especially focusing on risk communication, radiation education to public, and stakeholder involvements into decision making in local governments on radiation protection, relating to the accident.

• Environmental Analysis Health and Toxicology
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• v.20 no.4 s.51
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• pp.365-374
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• 2005

Ambient air particulate matters are classified into two distinct modes in sire distribution, namely the coarse and fine particles. Correlation between high particulate concentration and adverse effect on human populations has long been recognized. However, the toxicology of these adverse efforts has not been clarified. We investigated the genotoxic effect of PM 2.5 collected from urban area in Seoul by comet assay (A549 cells), CBMN assay (CHO-K1 cells) and EROD-microbioassay (H4IIE cells). Results from in vitro micronucleus assay and comet assay showed that PM 2.5 samples collected from traffic area, residential area and indoor air induced chromosomal damage and DNA breakage in a non-cytotoxic dose. The complex mixture effect of these PM 2.5 extracts was quantified by EROD-microbioassay in terms of its bio-TEQ (biologiral -TCDD equivalent concentration) which was 70.87$\pm$28.07, 93.55$\pm$21.80 and 14.31 $\pm$ 1.10 ng/g-PM 2.5 in traffic area, residental area and indoor air samples, respectively. Conclusively, we suggested that PM 2.5 collected from traffic area and residential area contains CYPIA inducer and genotoxic materials.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.939-944
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• 2015

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a $^{252}Californium$ ($^{252}Cf$) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1-9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.

• Journal of Radiation Protection and Research
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• v.38 no.3
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• pp.143-148
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• 2013

Operational nuclear facilities such as nuclear power plants and particle accelerators show various neutron spectra according to the type of facilities and specific position. Necessities of neutron dose management and neutron monitoring for radiation protection of radiation workers in such a kind of facilities have continuously increased in recent years. Bonner sphere spectrometers are widely used for measurement of neutron spectra. Data on response function of neutron detector, default neutron spectra and count rates of Bonner sphere spectrometer are required to obtain unfolded neutron spectra in specific workplaces. In this study, we carried out measurement of neutron spectra produced in MC50 cyclotron using Bonner sphere spectrometer with LiI scintillation detector. Additionally, we estimated quantitative data on neutron flux, mean neutron energy and ambient dose equivalent rate according to the incident proton energies and positions in MC50 cyclotron.