• Title/Summary/Keyword: 내부피폭선량평가

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Trends and Issues in Metabolism and Dosimetry for Tritium Intake (삼중수소 피폭방사선량 평가의 경향과 이슈에 대한 고찰)

  • Kim, Hee-Geun;Kong, Tae-Young;Jeong, Woo-Tae
    • Journal of Radiation Protection and Research
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    • v.36 no.2
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    • pp.99-106
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    • 2011
  • Tritium is the one of the most important radionuclide for workers in nuclear power plants (NPPs) and the public, from the dosimetric point of view. Humans are likely to have internal radiation exposure by tritium inhalation. Radiation exposure by tritium accounts for approximately 7% and 60~90% of the total radiation exposure of NPP workers and the public during normal operation, respectively. Thus, many researches have been conducted to estimate the internal dose by tritium precisely in the world. In terms of tritium dosimetry, this paper provides the current status of research for tritium metabolism and dosimetry.

Development of a Monte Carlo Simulation Code (CALEFF) for Calibrating Thyroid Internal Dose Measurement and Detection Efficiency Calculation (갑상선 내부피폭선량 측정치 보정을 위한 몬테카를로 모의실험 코드 (CALEFF) 개발 및 검출효율 계산)

  • Ahn, Ki-Soo;Cho1, Hyo-Sung
    • Journal of radiological science and technology
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    • v.28 no.2
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    • pp.117-122
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    • 2005
  • According to the Para. 5 of Art 2 of the Korean Nuclear Safety Regulations, which was revised in 1999, internal dose assessment as well as external one should be performed by law for employees at a nuclear power plant from 2003, and their estimate errors should also be within 50%. Thus, more accurate internal dosimetry becomes important. Corresponding to such regulation revision, we are developing a more accurate thyroid-uptake internal dosimetric system and have developed a Monte Carlo simulation code, the so-called CALEFF, to calculate the detection efficiency of the dosimetric system. In this paper, we calculated detection efficiencies with various test conditions by using the CALEFF code and discussed their characteristics. We may use the detection efficiency calculated by the code in calibrating the thyroid internal dose from measured data.

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The Experience on Intake Estimation and Internal Dose Assessment by Inhalation of Iodine-131 at Korean Nuclear Power Plants (국내 원전에서 $^{131}I$ 내부 흡입 에 따른 섭취량 산정과 내부피폭 방사선량 평가 경험 몇 개선방향에 대한 연구)

  • Kim, Hee-Geun;Kong, Tae-Young
    • Journal of Radiation Protection and Research
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    • v.34 no.3
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    • pp.129-136
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    • 2009
  • During the maintenance period at Korean nuclear power plants, internal exposure of radiation workers occurred by the inhalation of $^{131}I$ released to the reactor building when primary system opened. The internal radioactivity of radiation workers contaminated by $^{131}I$ was measured using a whole body counter. Intake estimation and the calculation of committed effective dose were also conducted conforming to the guidance of internal dose assessments from publications of International Commission on Radiological Protection. Because the uptake and excretion of $^{131}I$ in a body occur quickly and $^{131}I$ is accumulated in the thyroid gland, the estimated intakes showed differences depending on the counting time after intake. In addition, since ICRP publications do not provide the intake retention fraction (IRF) for whole body of $^{131}I$, the IRF for thyroid was substitutionally used to calculate the intake and subsequently this caused more error in intake estimation. Thus, intake estimation and the calculation of committed effective dose were conducted by manual calculation. In this study, the IRF for whole body was also calculated newly and was verified. During this process, the estimated intake and committed effective dose were reviewed and compared using several computer codes for internal dosimetry.

Intercomparison Exercise on Internal Dose Assessment in Korea (국내 내부피폭방사선량 평가 상호비교)

  • Lee, Jong-Il;Kim, Jang-Lyul;Kim, Bong-Hwan
    • Journal of Radiation Protection and Research
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    • v.36 no.2
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    • pp.64-70
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    • 2011
  • The intercomparison exercise on internal dose assessment has been carried out for the purpose of the evaluation for harmonization of internal dosimetry between the nuclear-related institutes in Korea. The exercises of 9 items on internal dose assessment have been developed for the unknown internal dosimetric parameters such as the intake pathway, absorption type, AMAD, and intake time of a radionuclide. Solutions to these exercises were reported by 7 participants from 5 institutes. The range of the ratio between the individual values and the geometric mean value of the evaluated doses for the exercises was $5.75{\times}10^{-4}$ ~ 9.81. But without the extreme partial solution, the range of the ratio was 0.216 ~ 3.12.

Preliminary Study on the Internal Dosimetry Program for Carbon-14 at Korean CANDU Reactors (중수로원전에서 발생하는 $^{14}C$에 대한 내부피폭 선량평가 프로그램에 관한 예비 조사)

  • Kong T.Y.;Kim H.C.;Park G.;Hang D.W.;Lee G.J.;Lee S.K.;Park S.C.
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2005.11a
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    • pp.317-320
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    • 2005
  • More strict radioactive regulations are applied to Korean nuclear power plants (NPPs) since ICRP-60 recommendation for radiation protection and has been enforced since 2003. In particular. carbon-14 and tritium concentrations are significantly higher at CANDU reactors compared to PWR reactors and this increases the risk of internal radiation exposure to workers at CANDU NPPs. Thus, it is necessary to estimate the exact amount of internal radiation exposure to workers fur radiological protection at CANDU reactors. In this paper, the current dosimetry method for carbon-14 is analyzed for the establishment of internal dosimetry for carbon-14 at domestic NPPs.

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삼중수소 내부피폭에 관한 연구

  • 박문수;곽성우;강창순
    • Proceedings of the Korean Nuclear Society Conference
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    • 1995.05b
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    • pp.913-917
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    • 1995
  • 인체내로 흡입된 삼중수소에 의한 영향을 평가하기 위한 기존의 내부피폭 평가 모델들을 검토하고, 이를 사용하여 body water와 OBT에 의한 선량을 계산하였다. 또한 이 모델들의 단점들을 도출하고, 이를 보안하기 위한 방안을 제시하였다.

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Internal Dose Assessment of Worker by Radioactive Aerosol Generated During Mechanical Cutting of Radioactive Concrete (원전 방사성 콘크리트 기계적 절단의 방사성 에어로졸에 대한 작업자 내부피폭선량 평가)

  • Park, Jihye;Yang, Wonseok;Chae, Nakkyu;Lee, Minho;Choi, Sungyeol
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.18 no.2
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    • pp.157-167
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    • 2020
  • Removing radioactive concrete is crucial in the decommissioning of nuclear power plants. However, this process generates radioactive aerosols, exposing workers to radiation. Although large amounts of radioactive concrete are generated during decommissioning, studies on the internal exposure of workers to radioactive aerosols generated from the cutting of radioactive concrete are very limited. In this study, therefore, we calculate the internal radiation doses of workers exposed to radioactive aerosols during activities such as drilling and cutting of radioactive concrete, using previous research data. The electrical-mobility-equivalent diameter measured in a previous study was converted to aerodynamic diameter using the Newton-Raphson method. Furthermore, the specific activity of each nuclide in radioactive concrete 10 years after nuclear power plants are shut down was calculated using the ORIGEN code. Eventually, we calculated the committed effective dose for each nuclide using the IMBA software. The maximum effective dose of 152Eu constituted 83.09% of the total dose; moreover, the five highest-ranked elements (152Eu, 154Eu, 60Co, 239Pu, 55Fe) constituted 99.63%. Therefore, we postulate that these major elements could be measured first for rapid radiation exposure management of workers involved in decommissioning of nuclear power plants, even if all radioactive elements in concrete are not considered.

Assessment of Inhalation Dose Sensitivity by Physicochemical Properties of Airborne Particulates Containing Naturally Occurring Radioactive Materials (천연방사성물질을 함유한 공기 중 부유입자 흡입 시 입자의 물리화학적 특성에 따른 호흡방사선량 민감도 평가)

  • Kim, Si Young;Choi, Cheol Kyu;Park, Il;Kim, Yong Geon;Choi, Won Chul;Kim, Kwang Pyo
    • Journal of Radiation Protection and Research
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    • v.40 no.4
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    • pp.216-222
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    • 2015
  • Facilities processing raw materials containing naturally occurring radioactive materials (NORM) may give rise to enhanced radiation dose to workers due to chronic inhalation of airborne particulates. Internal radiation dose due to particulate inhalation varies depending on particulate properties, including size, shape, density, and absorption type. The objective of the present study was to assess inhalation dose sensitivity to physicochemical properties of airborne particulates. Committed effective doses to workers resulting from inhalation of airborne particulates were calculated based on International Commission on Radiological Protection 66 human respiratory tract model. Inhalation dose generally increased with decreasing particulate size. Committed effective doses due to inhalation of $0.01{\mu}m$ sized particulates were higher than doses due to $100{\mu}m$ sized particulates by factors of about 100 and 50 for $^{238}U$ and $^{230}Th$, respectively. Inhalation dose increased with decreasing shape factor. Shape factors of 1 and 2 resulted in dose difference by about 18 %. Inhalation dose increased with particulate mass density. Particulate mass densities of $11g{\cdot}cm^{-3}$ and $0.7g{\cdot}cm^{-3}$ resulted in dose difference by about 60 %. For $^{238}U$, inhalation doses were higher for absorption type of S, M, and F in that sequence. Committed effective dose for absorption type S of $^{238}U$ was about 9 times higher than dose for absorption F. For $^{230}Th$, inhalation doses were higher for absorption type of F, M, and S in that sequence. Committed effective dose for absorption type F of $^{230}Th$ was about 16 times higher than dose for absorption S. Consequently, use of default values for particulate properties without consideration of site specific physiochemical properties may potentially skew radiation dose estimates to unrealistic values up to 1-2 orders of magnitude. For this reason, it is highly recommended to consider site specific working materials and conditions and use the site specific particulate properties to accurately access radiation dose to workers at NORM processing facilities.