• 한국지하수토양환경학회지:지하수토양환경
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• 제23권4호
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• pp.1-15
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• 2018

Despite of safety issues related to radiological hazards, 31 countries around the world are operating more than 450 nuclear power plants (NPPs). To operate NPPs safely, safety regulations from radiation protection organizations were developed and adopted in many countries. However, many cases of radionuclide releases at foreign NPPs have been reported. Almost all commercial NPPs routinely release radioactive materials to the surrounding environments as liquid and gas phases under control. These releases are called 'planned releases' which are planned, regularly monitored, and well documented. Meanwhile, the releases focused in this review, called 'unplanned releases', are neither planned nor monitored by regulatory and/or protection organizations. NPPs are generally composed of various structures, systems and components (SSCs) for safety. Among them, the SSCs near reactors are closely related to safety of NPPs, and typically fabricated to comply with stringent requirements. However, some non-safety related SSCs such as underground pipes may be constructed only according to commercial standards, causing the leakage of radioactive fluids usually containing tritium ($^3H$). This paper discusses SSCs of NPPs and introduces several cases of unplanned releases at foreign NPPs. The current regulation on the environmental radiological surveillance and assessment around the NPPs in South Korea are also examined.

• Journal of Radiation Protection and Research
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• 제45권2호
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• pp.81-87
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• 2020

Background: Kori unit #1 is permanently shut down after a 40-year lifetime. The Nuclear Safety and Security Commission recommends establishing initial decommissioning plans for all nuclear and radwaste treatment facilities. Therefore, the Korea Atomic Energy Research Institute (KAERI) must establish an initial and final decommissioning plan for radwaste-treatment facilities. Radiation safety assessment, which constitutes one chapter of the decommissioning plan, is important for establishing a decommissioning schedule, a strategy, and cost. It is also a critical issue for the government and public to understand. Materials and Methods: This study provides a method for assessing external radiation dose to workers during decommissioning. An external dose is calculated following each exposure scenario, decommissioning strategy, and working schedule. In this study, exposure dose is evaluated using the deterministic method. Physical characterization of the facility is obtained by both direct measurement and analysis of the drawings, and radiological characterization is analyzed using the annual report of KAERI, which measures the ambient dose every month. Results and Discussion: External doses are calculated at each stage of a decommissioning strategy and found to increase with each successive stage. The maximum external dose was evaluated to be 397.06 man-mSv when working in liquid-waste storage. To satisfy the regulations, working period and manpower must be managed. In this study, average and cumulative exposure doses were calculated for three cases, and the average exposure dose was found to be about 17 mSv/yr in all the cases. Conclusion: For the three cases presented, the average exposure dose is well below the annual maximum effective dose restriction imposed by the international and domestic regulations. Working period and manpower greatly affect the cost and entire decommissioning plan; hence, the chosen option must take account of these factors with due consideration of worker safety.

• Journal of Radiation Protection and Research
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• 제42권1호
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• pp.33-41
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• 2017

Background: Phosphate rock and its by-product are widely used in various industries to produce phosphoric acid, gypsum, gypsum board, and fertilizer. Owing to its high level of natural radioactive nuclides (e.g., $^{238}U$ and $^{226}Ra$), the radiological safety of workers who work with phosphate rock should be systematically managed. In this study, $^{238}U$, $^{232}Th$, $^{226}Ra$, and $^{40}K$ levels were measured to analyze the transport characteristics of these radionuclides in the production cycle of phosphate rock. Materials and Methods: Energy dispersive X-ray fluorescence and gamma spectrometry were used to determine the activity of $^{238}U$, $^{232}Th$, $^{226}Ra$, and $^{40}K$. To evaluate the extent of secular disequilibrium, the analytical results were compared using statistical methods. Finally, the distribution of radioactivity across different stages of the phosphate rock production cycle was evaluated. Results and Discussion: The concentration ratios of $^{226}Ra$ and $^{238}U$ in phosphate rock were close to 1.0, while those found in gypsum and fertilizer were extremely different, reflecting disequilibrium after the chemical reaction process. The nuclide with the highest activity level in the production cycle of phosphate rock was $^{40}K$, and the median $^{40}K$ activity was $8.972Bq{\cdot}g^{-1}$ and $1.496Bq{\cdot}g^{-1}$, respectively. For the $^{238}U$ series, the activity of $^{238}U$ and $^{226}Ra$ was greatest in phosphate rock, and the distribution of activity values clearly showed the transport characteristics of the radionuclides, both for the byproducts of the decay sequences and for their final products. Conclusion: Although the activity of $^{40}K$ in k-related fertilizer was relatively high, it made a relatively low contribution to the total radiological effect. However, the activity levels of $^{226}Ra$ and $^{238}U$ in phosphate rock were found to be relatively high, near the upper end of the acceptable limits. Therefore, it is necessary to systematically manage the radiological safety of workers engaged in phosphate rock processing.

• Journal of Radiation Protection and Research
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• 제48권3호
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• pp.162-166
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• 2023

Background: Assessment of the radiation doses to which workers are exposed can differ depending on the placement of dosimeters on the body. In addition, it is affected by whether the placement is under or over a shielding apron. This study aimed to evaluate the actual positioning of personal dosimeters on the body, with or without shielding aprons, among radiation workers in Korea. Materials and Methods: We analyzed the survey data, which included demographic characteristics, such as sex, age, occupation, work history, and placement of the personal dosimeter being worn, from a cohort study of Korean radiation workers. We assessed the use of personal dosimeters among workers, stratified by sex, age, working period, starting year of work, and occupation. Results and Discussion: Overall, high compliance (89.1% to 99.0%) with the wearing of dosimeters on the chest was observed regardless of workers' characteristics, such as age, sex, occupation, and work history. However, the placement of dosimeters, either under or over the shielding aprons, was inconsistent. Overall, 40.1% of workers wore dosimeters under their aprons, while the others wore dosimeters over their aprons. This inconsistency indicates that radiation doses are possibly measured differently under the same exposure conditions solely owing to variations in the placement of worn dosimeters. Conclusion: Although a lack of uniformity in dosimeter placement when wearing a shielding apron may not cause serious harm in radiation dose management for workers, the development of detailed guidelines for dosimeter placement may improve the accuracy of dose assessment.

• Journal of Radiation Protection and Research
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• 제43권4호
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• pp.143-153
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• 2018

Background: The determination of the amount of radionuclides and internal dose for the worker who may have intake of radionuclides results in a variation due to uncertainty of measurement data and ingestion information. As a result of this, it is possible that for the same internal exposure scenario assessors could make considerably different estimation of internal dose. In order to reduce this difference, internal exposure scenarios for nuclear facilities were developed, and intercomparison were made to determine the harmonization of dose assessment results among the assessors. Materials and Methods: Seven cases on internal exposures incidents that have occurred or may occur were prepared by referring to the intercomparison excercise scenario that NRC and IAEA have carried out. Based on this, 16 nuclear facilities concerned with internal exposure in Korea were asked to evaluate the scenarios. Each result was statistically determined according to the harmonization discrimination criteria developed by IDEAS/IAEA. Results and Discussion: The results were evaluated as having no outliers in all 7 cases. However, the distribution of the results was spread by various causes. They can be divided into two wide categories. The first one is the distribution of the results according to the assumption of the intake factors and the evaluation factors. The second one is distribution due to misapplication of calculation method and factors related to internal exposure. Conclusion: In order to satisfy the harmonization criteria and accuracy of the internal exposure dose evaluation, it is necessary that exact guidelines should be set on low dose, and various intercomparison cases also be needed including high dose exposure as well as the specialized education. The aim of the blind test is to make harmonization evaluation, but it will also contribute to securing the expertise and high quality of dose evaluation data through the discussion among the participants.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2379-2386
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• 2020

This paper presents the results of the first intercomparison exercise performed by the Korea retrospective dosimetry (KREDOS) working group using electron paramagnetic resonance (EPR) spectroscopy. The intercomparison employed the alanine dosimeter, which is commonly used as the standard dosimeter in EPR methods. Four laboratories participated in the dose assessment of blind samples, and one laboratory carried out irradiation of blind samples. Two types of alanine dosimeters (Bruker and Magnettech) with different geometries were used. Both dosimeters were blindly irradiated at three dose levels (0.60, 2.70, and 8.00 Gy) and four samples per dose were distributed to the participating laboratories. Assessments of blind doses by the laboratories were performed using their own measurement protocols. One laboratory did not participate in the measurements of Magnettech alanine dosimeter samples. Intercomparison results were analyzed by calculating the relative bias, E_n value, and z-score. The results reported by participating laboratories were overall satisfactory for doses of 2.70 and 8.00 Gy but were considerably overestimated with a relative bias range of 10-95% for 0.60 Gy, which is lower than the minimum detectable dose (MDD) of the alanine dosimeter. After the first intercomparison, participating laboratories are working to improve their alanine-EPR dosimetry systems through continuous meetings and are preparing a second intercomparison exercise for other materials.

• Nuclear Engineering and Technology
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• 제40권4호
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• pp.277-284
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• 2008

Recently, the long-term operation of a nuclear power plant beyond its licensed term has become a worldwide trend as long as the safety of the plant is maintained in the extended period. Kori Unit 1, the oldest PWR in Korea, is the foremost example of this type of long-term operation in Korea. Comprehensive technical evaluation of the long-term operation of this plant was completed to confirm the overall safety of the plant. The technical evaluation included a review of PSR results, an assessment on aging management programs and time limited aging analyses, and a statement of radiological impact on the environment. Based on all of the results of the technical evaluation activities, Kori Unit 1 was approved to operate for an additional 10 years beyond its original design life of 30 years.

• 대한방사선기술학회지:방사선기술과학
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• 제44권5호
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• pp.495-502
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• 2021

This study aims to provide basic data for elderly health insurance policy and medical radiation safety management by analyzing the general radiography usage and exposure dose of the elderly in Korea. The effective dose for each general radiography was calculated using the ALARA-GR program for 260 general radiography codes selected from 'National Health Insurance Care Benefit Cost'. The usage of general radiography was analyzed in the 2016 elderly patient data of the Health Insurance Review and Assessment Service, and the effective dose for each general radiography was applied. The general radiography usage and exposure dose per person aged 65 years and over was 6.47 cases and 0.56 mSv. Females showed higher value than males as 7.15 cases and 0.66 mSv(p<.001). By age, those between 75 and 79 showed the highest number as 6.97 cases and 0.62 mSv(p<.001). Those who were supported by Medical Aid showed higher value than those who were insured by National Health Insurance as 8.82 cases and 0.76 mSv(p<.001). In addition, the ratio by radiography was in the order of Chest 20.85%, Knee Joint 15.58%, and L-spine 14.67%, and the exposure dose was L-spine 29.40%, Chest 15.82%, Abdomen 7.97%, and Entire Spine 7.20%. General radiography, which is widely used due to the high frequency of diseases in the elderly population should be taken into consideration when establishing health insurance policies. In addition, it is necessary to check whether the general radiography with high exposure dose is performed as a routine examination without considering medical necessity.

• 한국방사선학회논문지
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• 제12권6호
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• pp.745-753
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• 2018

방사선이 다양한 분야에 사용됨에 따라 방사선작업종사자의 수가 증가하고 있으며, 이에 따라 종사자의 안전에 대한 우려도 높아지고 있다. 국내에서는 KISOE 시스템, 발주자보고 등을 통해 종사자의 안전 확보에 주력하고 있다. 선행연구에서는 종사자 및 업체의 안전 확보를 위한 위험성 평가에 피폭선량과 더불어 이외의 항목에 대한 추가적 평가가 효과적일 것이라고 판단하여 평가를 위한 항목들 및 방사형 척도분석 모델을 개발하였다. 이에 본 연구에서는 2016년 방사선투과검사업체의 자료를 방사형척도분석 모델에 적용하여 실제 업체의 위험성 평가를 수행하였다. 또한 위험성이 낮을 것으로 예상되는 업체 2곳과 위험성이 높을 것으로 예상되는 업체 2곳을 선정하여 위험성 분석을 수행하였다. 분석 결과에서 예상과 동일한 결과를 얻어 모델의 타당성을 확인할 수 있었다. 전체 56개 업체의 위험성 평가가 수행되었고 업체별 문제점에 따른 개선 및 점검사항을 예측하였다. 본 연구 결과를 방사선투과검사업체 및 규제기관에서 자체 평가 및 규제 기준 등으로 활용할 수 있을 것으로 예상된다.

• 방사성폐기물학회지
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• 제15권2호
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• pp.161-172
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• 2017

중 저준위 방사성폐기물 처분시설의 운영 중 사고로 인한 방사선적 영향을 평가하기 위해서는 운영 중 발생 가능한 사고에 대한 타당성이 입증되어야 한다. 본 논문에서는 처분시설의 운영 중 사고분석 체계를 처분시설의 구성요소에 대한 안전기능분석, 잠재위험요소분석, 위험도분석, 그리고 향후 조치대안으로 사고평가체계를 개선하였다. 이를 위하여 위험도분석에 필요한 설계대안과 관리대안을 추가하여 설계-운영-평가가 연계되도록 하였다. 또한 운영 중 사고의 발생확률과 평가결과의 심각성에 따라 운영중 사고에 대한 분류기준을 제안하여 처분시설 운영 중 대표 사고시나리오에 대한 정당성을 확보하였다. 본 논문의 개선된 평가체계를 우리나라의 2단계 중 저준위 방사성폐기물 표층처분시설에 대한 처분시설 운영 중 사고분석의 사례에 대해 적용하였다.