• 한국지하수토양환경학회지:지하수토양환경
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• 제20권6호
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• pp.1-7
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• 2015

The most critical health effect of lead exposure is the neurodevelopmental effect to children caused by the increased blood lead level. Therefore, the endpoint of the risk assessment for lead-contaminated sites should be set at the blood lead level of children. In foreign countries, the risk assessment for lead-contaminated sites is conducted by estimating the increased blood lead level of children via oral intake and/or inhalation (United States Environmental Protection Agency, USEPA), or by comparing the estimated oral dose to the threshold oral dose of lead, which is derived from the permissible blood lead level of children (Dutch National Institute for Public Health and the Environment, RIVM). For the risk assessment, USEPA employs Integrated-Exposure-Uptake-Biokinetic (IEUBK) Model to check whether the estimated portion of children whose blood lead level exceeds 10 µg/dL, threshold blood lead level determined by USEPA, is higher than 5%, while Dutch RIVM compares the estimated oral dose of lead to the threshold oral dose (2.8 µg/kg-day), which is derived from the permissible blood lead level of children. In Korea, like The Netherlands, risk assessment for lead-contaminated sites is conducted by comparing the estimated oral dose to the threshold oral dose; however, because the threshold oral dose listed in Korean risk assessment guidance is an unidentified value, it is recommended to revise the existing threshold oral dose described in Korean risk assessment guidance. And, if significant lead exposure via inhalation is suspected, it is useful to employ IEUBK Model to derive the risk posed via multimedia exposure (i.e., both oral ingestion and inhalation).

• 방사선산업학회지
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• 제17권3호
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• pp.265-273
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• 2023

To effectively and safely manage the radiation exposure to nuclear power plant (NPP) workers in accidents, major overseas NPP operators such as the United States, Germany, and France have developed and applied realistic 3D model radiation dose assessment software for workers. Continuous research and development have recently been conducted, such as performing NPP accident management using 3D-VR based on As Low As Reasonably Achievable (ALARA) planning tool. In line with this global trend, it is also required to secure technology to manage radiation exposure of workers in Korea efficiently. Therefore, in this paper, it is described the application method and assessment results of radiation exposure scenarios for workers in response to accidents assessment technology, which is one of the fundamental technologies for constructing a realistic platform to be utilized for radiation exposure prediction, diagnosis, management, and training simulations following accidents. First, the post-accident sampling after the Loss of Coolant Accident(LOCA) was selected as the accident and response scenario, and the assessment area related to this work was established. Subsequently, the structures within the assessment area were modeled using MCNP, and the radiation source of the equipment was inputted. Based on this, the radiation dose distribution in the assessment area was assessed. Afterward, considering the three principles of external radiation protection (time, distance, and shielding) detailed work scenarios were developed by varying the number of workers, the presence or absence of a shield, and the location of the shield. The radiation exposure doses received by workers were compared and analyzed for each scenario, and based on the results, the optimal accident response scenario was derived. The results of this study plan to be utilized as a fundamental technology to ensure the safety of workers through simulations targeting various reactor types and accident response scenarios in the future. Furthermore, it is expected to secure the possibility of developing a data-based ALARA decision support system for predicting radiation exposure dose at NPP sites.

• 한국식품위생안전성학회지
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• 제14권4호
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• pp.319-326
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• 1999

최근 국내에서는 Escherichia coli O157:H7, Listeria monocytogenes등의 미생물에 대한 건강위해성이 강조되면서 미생물 위해성평가에 대한 필요성이 제기되고 있고, U.S.FPA, U.S.DA, FAO/WHO를 비롯한 국제기구 및 외국 유관기관들에서도 미생물 위해성평가 방법을 식품관리에 활용하고자 방법론 연구에 주력하고 있다. 미생물 위험성은 화학물질과 달리 인체건강에 대한 영향이 즉각적이고, 심각하게 나타나 정량적인 위해성평가가 용이하지 않고 유해화학물질과는 다른 평가방법이 요구된다. 식품중 미생물의 위해성평가는 크게 4단계로 구분되는데, 미생물관련질환 추세파악 및 미생물 관련질병에 관한 역학조사 등을 활용하는 위험성확인 단계와 실제 식품원료, 식품가공, 수송, 포장단계 중 식품의 물리적, 화학적 조건에 따른 미생물 변화를 고려하여 식품중 미생물에 대한 노출을 정량화하는 노출평가 단계, 미생물의 용량에 따른 질병발생에 근거하여 용량-반응관계를 규명하는 용량-반응평가 단계, 규명된 모델을 활용하여 모든 평가결과를 통합함으로 위해 도치 예측과 불확실성 분석 등을 수행하는 위해도결정단계로 구성되어 있다. 미생물 용량-반응평가는 크게 비역치(Nonthreshold)와 역치(Threshold) 평가 방법론으로 구분되는데, 비역치 평가방법론은 단일 병원균이 감염을 일으킬 수 있다는 것과 감염을 일으킬 수 있는 확률이 독립적이라는 가정을 전제로 하고, 역치평가방법론은 미생물이 감염을 일으키기 위해서 각기 개별 역치가 존재하는데 어느 정도의 미생물수가 모여 서로 작용해야 독성유발물질을 만들어 낸다는 가정을 전제로 한다. 현재 받아들여지고 있는 비역치 모델로는 Exponential, Beta-poisson, Gompertz, Gamma-weibull 모델 등이 있으며, 역치모델로는 Log-normal, Log-logistic모델 등이 있다. 본 연구에서는 인체 volunteer자료를 활용하여 용량-반응자료를 입력하고 용량-반응자료를 토대로 적합한 수학적 모델을 찾아내어, 선별한 모델의 적합도 검정을 실시하는 방법론 연구를 실시하였으며, 노출평가 자료와 용량-반응평가 결과를 연계하여 위해도를 결정하는 과정에 대해 연구하였다 이 밖에도 모델(Food MicroModel)을 이용하여 식품의 염도, 수분활성도, 온도, pH등의 조건에 따른 미생물의 성장률, 사멸률 등 변화를 예측할 수 있는 방법론 연구를 통해 식품의 최적 보관 조건등을 찾아내는 방법을 습득하였다. 미생물 위해성평가는 외국에서도 아직 초기 연구단계에 있으며 현재로서 사후조사자료인 역학자료보다 건강한 성인남자를 대상으로 한 volunteer 자료를 우선적으로 활용하고 있으나 노약자나 민감그룹에 대한 실험은 현실적으로 불가능하여 동물실험을 이용한 평가방법을 연구중에 있다. 추후 연구방향으로는 국내 volunteer들을 대상으로 한 미생물별 용량-반응결과를 토대로 population sensitivity를 비교할 수 있는 기초자료를 확보함으로써 미생물에 대한 인구집단의 반응 민감성 차이를 비교하고 시료채취 후 즉각적인 실험실적 분석이 가능토록하여 정확한 인체노출평가를 수행함으로써 미생물 위해성평가방법론을 식품미생물관리에 적용하는 것이다.

• 방사선산업학회지
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• 제17권1호
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• pp.83-92
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• 2023

When assessing the risk of radioactive wastes transportation on land, computer codes such as RADTRAN and RISKIND are used as deterministic methods. Transportation risk assessment using the deterministic method requires a relatively short assessment time. On the other hand, transportation risk assessment using the probabilistic method requires a relatively long assessment time, but produces more reliable results. Therefore, a study is needed to evaluate the exposure dose using a deterministic method that can be evaluated relatively quickly, and to compare and analyze the exposure dose result using a probabilistic method. The purpose of this study is to evaluate the exposure dose during transportation of radioactive wastes using deterministic and probabilistic methods, and to compare and analyze them. For this purpose, the main exposure factors were selected and various exposure situations were set. The distance between the radioactive waste and the receptor, the size of the package, and the speed of vehicle were selected as the main exposure factors. The exposure situation was largely divided into when the radioactive wastes were stationary and when they were passing. And the dose (rate) model of the deterministic overland transportation risk assessment computer code was analyzed. Finally, the deterministic method of the RADTRAN computer code and the RISKIND computer code and the probabilistic method of the MCNP 6 computer code were used to evaluate the exposure dose in various exposure situations during transportation of radioactive wastes. Then we compared and analyzed them. As a result of the evaluation, the tendency of the exposure dose (rate) was similar when the radioactive wastes were stationary and passing. For the same situation, the evaluation results of the RADTRAN computer code were generally more conservative than the results of the RISKIND computer code and the MCNP 6 computer code. The evaluation results of the RISKIND computer code and the MCNP 6 computer code were relatively similar. The results of this study are expected to be used as basic data for establishing the radioactive wastes transportation risk assessment system in Korea in the future.

• 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.

• 한국환경성돌연변이발암원학회지
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• 제19권1호
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• pp.39-45
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• 1999

Human Carcinogenic Potency (HCP) can be estimated based on human daily exposure dose to carcinogen (Dh), body weight (Wh), 10% tumorigenic dose (TD10), and slope factor at TD10 (Q10) from 2-yr bioassay data. This approach is more relevant to humans generally exposed to low doses of carcinogens and can reduce more of extrapolation errors from high dose in animal experiments to low dose in humans than HERP (human exposure dose/rodent potency dose) proposed by Ames et al. (Science, 236, 271-280, 1987). TD50 and HERP have been routinely used to compare rodent carcinogenic potency and human carcinogenic potency, but those approaches have had limitations in extrapolation of high dose to low dose in humans. The advantages of HCP are to estimate human exposure dose (Dh) by human monitoring instead of environmental monitoring, to consider slope factor (Q10) which reflects the tendency of curve at low dose, and to use TD10 which represents much lower dose thant TD50 or HERP. HCP will be a useful parameter for the estimation of human carcinogenic potency in risk assessment and management of carcinogens.

• 방사선산업학회지
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• 제18권1호
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• pp.35-42
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• 2024

Demand for RW transportation is expected to increase due to the continuous generation of RW from nuclear power plants and facilities, decommissioning of plants, and saturation of spent fuel temporary storage facilities. The locational aspect of plants and radiation protection optimization for the public have led to an increasing demand for maritime transportation, necessitating to apprehend the overseas and domestic current status. Given the potential long-term radiological impact on the public in the event of a sinking accident, a pre-transportation exposure assessment is necessary. The objective of this study is to investigate the overseas and domestic RW maritime transportation current status and overseas dose assessment cases for the public in sinking accident. Selected countries, including Japan, UK, Sweden, and Korea, were examined for transport cases, Japan and the U.S were chosen for dose assessment case in sinking accidents. As a result of the maritime transportation case analysis, it was performed between nuclear power plants and reprocessing facilities, from plants to disposal or intermediate storage facilities. HLW and MOX fuel were transported using INF 3 shipments, and all transports were performed low speed of 13 kn or less. As a result of the dose assessment for the public in sinking accident, japan conducted an assessment for the sinking of spent fuel and vitrified HLW, and the U.S conducted for the sinking of spent fuel. Both countries considered external exposure through swimming and working at seashore, and internal exposure through seafood ingestion as exposure pathway. Additionally, Japan considered external exposure through working on board and fishing, and the U.S considered internal exposure through spray inhalation and desalinized water and salt ingestion. Internal exposure through seafood ingestion had the largest dose contribution. The average public exposure dose was 20 years after the sinking, 0.04 mSv yr^-1 for spent fuel and 5 years after the sinking, 0.03 mSv yr^-1 for vitrified HLW in Japan. In the U.S, it was 1.81 mSv yr^-1 5 years after the sinking of spent fuel. The results of this study will be used as fundamental data for maritime transportation of domestic RW in the future.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.424-435
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• 2016

Background: The International Commission on Radiological Protection (ICRP) recommendations and the Federal Guidance Report (FGR) published by the U.S. Environmental Protection Agency (EPA) have been widely applied worldwide in the fields of radiation protection and dose assessment. The dose conversion coefficients of the ICRP and FGR are widely used for assessing exposure doses. However, before the coefficients are used, the user must thoroughly understand the derivation process of the coefficients to ensure that they are used appropriately in the evaluation. Materials and Methods: The ICRP provides recommendations to regulatory and advisory agencies, mainly in the form of guidance on the fundamental principles on which appropriate radiological protection can be based. The FGR provides federal and state agencies with technical information to assist their implementation of radiation protection programs for the U.S. population. The system of radiation dose assessment and dose conversion coefficients in the ICRP and FGR is reviewed in this study. Results and Discussion: A thorough understanding of their background is essential for the proper use of dose conversion coefficients. The FGR dose assessment system was strongly influenced by the ICRP and the U.S. National Council on Radiation Protection and Measurements (NCRP), and is hence consistent with those recommendations. Moreover, the ICRP and FGR both used the scientific data reported by Biological Effects of Ionizing Radiation (BEIR) and United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) as their primary source of information. The difference between the ICRP and FGR lies in the fact that the ICRP utilized information regarding a population of diverse races, whereas the FGR utilized data on the American population, as its goal was to provide guidelines for radiological protection in the US. Conclusion: The contents of this study are expected to be utilized as basic research material in the areas of radiation protection and 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.

• 방사성폐기물학회지
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• 제19권1호
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• pp.123-132
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• 2021

Radiological impact analyses were carried out for a near-surface radioactive waste repository at Gyeongju in South Korea. The RESRAD-ONSITE code was applied for the estimation of maximum exposure doses by considering various exposure pathways based on a land area of 2,500 ㎡ with a 0.15 m thick contamination zone. Typical influencing input parameters such as shield depth, shield materials' density, and shield erosion rate were examined for a sensitivity analysis. Then both residential farmer and industrial worker scenarios were used for the estimation of maximum exposure doses depending on exposure duration. The radiation dose evaluation results showed that 60Co, 137Cs, and 63Ni were major contributors to the total exposure dose compared with other radionuclides. Furthermore, the total exposure dose from ingestion (plant, meat, and milk) of the contaminated plants was more significant than those assessed for inhalation, with maximum values of 5.5×10-4 mSv·yr-1 for the plant ingestion. Thus the results of this study can be applied for determining near-surface radioactive waste repository conditions and providing quantitative analysis methods using RESRAD-ONSITE code for the safety assessment of disposing radioactive materials including decommissioning wastes to protect human health and the environment.