• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.255-263
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• 2001

An Orphan Sources Control program controlled by a web based information system in Korea has been developed to satisfy the national demand on a total management of and integrated radiation safety. There are, currently, three approaches going on to control and manage the orphan sources in Korea. First, Korean regulatory authority has been conducting scrutinizing investigation on and thoroughly monitoring the possession of unlicensed radioactive sources from the late 1990s. Second, the regulatory authority will fully operate an information system on radiation safety to effectively trace and monitor all radioactive sources in the country by the mid 2001. Finally, the regulatory authority strongly advises steel mill companies to closely scrutinize and inspect the scrap metals when they attempt to reutilize metals to prevent from being contaminated by uncontrolled sources through the scrap monitoring systems.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1153-1164
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• 2024

In recent years, unmanned ground vehicles (UGVs) have been used to search for lost or stolen radioactive sources to avoid radiation exposure for operators. To achieve autonomous localization of radioactive sources, the UGVs must have the ability to automatically determine the next radiation measurement location instead of following a predefined path. Also, the radiation field of radioactive sources has to be reconstructed or inverted utilizing discrete measurements to obtain the radiation intensity distribution in the area of interest. In this study, we propose an effective source localization framework and method, in which UGVs are able to autonomously explore in the radiation area to determine the location of radioactive sources through an iterative process: path planning, radiation field reconstruction and estimation of source location. In the search process, the next radiation measurement point of the UGVs is fully predicted by the design path planning algorithm. After obtaining the measurement points and their radiation measurements, the radiation field of radioactive sources is reconstructed by the Gaussian process regression (GPR) model based on machine learning method. Based on the reconstructed radiation field, the locations of radioactive sources can be determined by the peak analysis method. The proposed method is verified through extensive simulation experiments, and the real source localization experiment on a Cs-137 point source shows that the proposed method can accurately locate the radioactive source with an error of approximately 0.30 m. The experimental results reveal the important practicality of our proposed method for source autonomous localization tasks.

• Journal of Radiation Protection and Research
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• 제12권2호
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• pp.37-50
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• 1987

The exposure of the population in the United States to ionizing radiation has recently been evaluated by the National Council on Radiation Protection and Measurements (NCRP). This was done by constituting six organizational groups to address various phases of the work and the results of this work are summarized in this article. The article is based on the report, by the same title, which is scheduled for publication by the NCRP in September, 1987. The six organizational groups are titled Radiation Exposure from Consumer Products, Natural Background Radiation, Radiation Associated with Medical Examinations, Radiation Received by Radiation Employees, Public Exposure from Nuclear Power, and Exposure from Miscellaneous Environmental Sources. These titles are descriptive of the subject areas covered by each of these separate groups. The data evaluated are for the years 1977-1984 with the majority of the data being for the period 1980-1982. Summary information is presented and discussed for the number of people exposed to given sources, the effective dose equivalent, the average effective dose equivalent to the U.S. population, and the genetically significant dose equivalent. The average annual effective dose equivalent from all sources to the U.S. population is approximately 3.6 mSv (360 mrem). Exposures to natural sources make the largest contribution to this total. Radon and radon decay products contribute 2.0 mSv (200 mrem) whereas the other naturally occurring radionuclides contribute 1.0 mSv (100 mrem). Among man-made or enhanced sources, medical exposures make the largest additional contributions, namely 0.39 mSv (39 mrem) for diagnosis and 0.14 mSv (14 mrem) for nuclear medicine. It was not possible to evaluate exposures for therapy. Most of the other sources of population exposure, including nuclear power and consumer products, are minor. A possible exception would be the use of tobacco products. These exposures are discussed in relation to a negligible individual risk level of $10{\mu}Sv/y$ (1 mrem/y). The NCRP considers exposures below the negligible individual risk level as trivial and as such should be dismissed.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.149-154
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• 2001

Calculational models to evaluate radiation doses resulting from the medical use of high energy beta-ray sources are presented. The radioactive sources considered are Sr-90/Y-90 used as ophthalmic applicator, Re-188 used for treating restenosis of coronary artery, and Ho-166 used for treating hepatic tumors. Typical therapeutic situations which might induce relatively high radiation doses the medical person involved were considered to compute by using MCNP-4C Monte Carlo code the radiation doses. The calculation results suggest that for all of the cases considered, the evaluated radiation doses are negligible compared to the dose limits. It is also found that the effect of Bremsstrahlung radiations on the total dose is insignificant, and hence the conventional lead gown is also effective in shielding beta-rays.

• 방사선산업학회지
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• 제17권3호
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• pp.275-282
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• 2023

Workers in nuclear power plants are likely to be exposed to radiation from various geometrical sources. In order to evaluate the exposure level, the point-kernel method can be utilized. In order to perform a dose assessment based on this method, the radiation source should be divided into point sources, and the number of divisions should be set by the evaluator. However, for the general public, there may be difficulties in selecting the appropriate number of divisions and performing an evaluation. Therefore, the purpose of this study is to develop an algorithm for dose assessment for arbitrary shaped sources based on the point-kernel method. For this purpose, the point-kernel method was analyzed and the main factors for the dose assessment were selected. Subsequently, based on the analyzed methodology, a dose assessment algorithm for arbitrary shaped sources was developed. Lastly, the developed algorithm was verified using Microshield. The dose assessment procedure of the developed algorithm consisted of 1) boundary space setting step, 2) source grid division step, 3) the set of point sources generation step, and 4) dose assessment step. In the boundary space setting step, the boundaries of the space occupied by the sources are set. In the grid division step, the boundary space is divided into several grids. In the set of point sources generation step, the coordinates of the point sources are set by considering the proportion of sources occupying each grid. Finally, in the dose assessment step, the results of the dose assessments for each point source are summed up to derive the dose rate. In order to verify the developed algorithm, the exposure scenario was established based on the standard exposure scenario presented by the American National Standards Institute. The results of the evaluation with the developed algorithm and Microshield were compare. The results of the evaluation with the developed algorithm showed a range of 1.99×10^-1~9.74×10^-1 μSv hr^-1, depending on the distance and the error between the results of the developed algorithm and Microshield was about 0.48~6.93%. The error was attributed to the difference in the number of point sources and point source distribution between the developed algorithm and the Microshield. The results of this study can be utilized for external exposure radiation dose assessments based on the point-kernel method.

• 방사선산업학회지
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• 제8권2호
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• pp.83-88
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• 2014

Application of X-ray irradiation of dried condiments was studied using commercially prepared dried garlic, onion and welsh onion flakes as model samples. Total load of aerobic microbes (TAM), color differences, and generation of off-flavor were quantified for samples individually irradiated with gamma rays, electrons, or X-rays. TAM load was decreased by irradiation in a dose-dependent manner. The three types of radiation did not differ in the extent of TAM reduction (P>0.05). The samples did not differ in color. Off-flavor was detected from 6 kGy-irradiated samples, regardless of radiation sources. The results indicated that X-ray irradiation could be used for irradiation of dried condiments with the same effects as gamma rays and an electron beam.

• Journal of Radiation Protection and Research
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• 제14권1호
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• pp.66-70
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• 1989

방사성 중성자선원은 일상적 시험에 있어 표준 중성자 방사선장을 형성하는데 적합하다. 방사선 방어상의 목적으로 사용되는 중성자 측정기기의 교정을 위한 기준 방사선이 ISO TC-85에서 제의되었다. 한국표준연구소 방사선연구실에는 ISO TC-85의 추천사항에 준하여 개인용 중성자 선량계를 교정하기 위하여 $^{252}Cf$와 $^{241}Am-Be$ 선원을 이용한 표준조사시설을 설립하였다. 본 연구에서는 중성자 산란과 선원 비등방성에 연관된 교정상의 보정인자들을 실험에 의하여 결정하였다.

• 한국식품저장유통학회지
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• 제21권3호
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• pp.315-320
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• 2014

본 연구에서는 후춧가루의 위생화를 위해 감마선, 전자선 및 X-선을 조사한 후 이에 따른 미생물학적 및 관능적 품질변화를 평가하였다. 후춧가루에 감마선, 전자선 및 X-선을 2, 4, 6, 8 및 10 kGy로 각각 조사한 후 미생물 변화와 관능품질 변화를 평가한 결과, 미생물의 경우 선종에 관계없이 흡수선량이 증가할수록 미생물 제어 효과가 증가하는 것으로 나타났으며, 선종별 D10값의 경우 2.24-2.75 kGy범위를 보였으며 방사선 선종에 따른 유의차가 없는 것으로 나타났다. 색도의 경우 명도, 적색도 및 황색도 모두 방사선 선종에 따른 변화가 없는 것으로 판단하였다. 관능적 품질 역시 방사선 선종에 관계없이 흡수선량에 따라 품질저하가 발생되는 것으로 나타났다. 이상의 결과를 종합하여 볼 때 방사선 선종에 따른 후춧가루의 품질변화 차이가 없는 것으로 판단하였으며, 이상의 결과는 X-선의 상업적 이용을 위한 기초 연구 자료로 활용될 수 있을 것으로 사료되었다.

• 방사선산업학회지
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• 제6권2호
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• pp.181-188
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• 2012

A mushroom mutant with increased cellulolytic activity was developed through radiation mutagenesis. The homogenized hypha suspension of Pleurotus florida was exposed to gamma radiation ($^{60}Co$, AECL) at the dose of $LD_{99}$ (0.51 kGy, $D_{10}$; 0.26 kGy). Among 16 mutants, Pf CM4 showed 17.24% more cellulolytic activity than the wild type (p<0.05). It was observed that Pf CM4 can utilize all kinds of carbon sources tested for their mycelia growth. Starch, xylan, and glucose favourably supported the radial mycelia extension. Yeast extract and $NH_4NO_3$ have been recorded as the best organic and inorganic nitrogen sources, respectively. Pf CM4 was found to grow significantly faster, even at high temperature ($30^{\circ}C$), than wild type (p<0.05), and the optimal pH was 5.5~6.5. This study reveals that the mutant Pf CM4 could be employed for the effective recycling of cellulosic wastes, in addition to mushroom farming.

• 분석과학
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• 제35권2호
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• pp.69-81
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• 2022

The method of measuring and classifying the energy category of neutrons directly using raw data acquired through a CZT detector is not satisfactory, in terms of accuracy and efficiency, because of its poor energy resolution and low measurement efficiency. Moreover, this method of measuring and analyzing the characteristics of low-energy or low-activity gamma-ray sources might be not accurate and efficient in the case of neutrons because of various factors, such as the noise of the CZT detector itself and the influence of environmental radiation. We have therefore developed an efficient method of analyzing radiation characteristics using a neutron and gamma-ray analysis algorithm for the rapid and clear identification of the type, energy, and radioactivity of gamma-ray sources as well as the detection and classification of the energy category (fast or thermal neutrons) of neutron sources, employing raw data acquired through a CZT detector. The neutron analysis algorithm is based on the fact that in the energy-spectrum channel of 558.6 keV emitted in the nuclear reaction ¹¹³Cd + ¹n → ¹¹⁴Cd + in the CZT detector, there is a notable difference in detection information between a CZT detector without a PE modulator and a CZT detector with a PE modulator, but there is no significant difference between the two detectors in other energy-spectrum channels. In addition, the gamma-ray analysis algorithm uses the difference in the detection information of the CZT detector between the unique characteristic energy-spectrum channel of a gamma-ray source and other channels. This efficient method of analyzing radiation characteristics is expected to be useful for the rapid radiation detection and accurate information collection on radiation sources, which are required to minimize radiation damage and manage accidents in national disaster situations, such as large-scale radioactivity leak accidents at nuclear power plants or nuclear material handling facilities.