• Journal of Radiation Protection and Research
• /
• v.23 no.4
• /
• pp.211-218
• /
• 1998

The distributional effects of radioactive materials on external gamma exposure have been analyzed. An approximate method for estimating external gamma dose given from an arbitrary distribution of radioactive material has been developed. The minimum gamma exposure given from a point source is shown at 0.07 MeV if the source to receptor distance is shorter than 10 m. But if the receptor to point source distance is longer than 20 m, gamma exposure rate increases monotonously according to the average gamma energy. For the analysis of the effects of volume source, we estimated the gamma dose given from different size of hemisphere in which radioactive materials are distributed uniformly. When the radius of hemisphere is longer than 40 m, external gamma dose rate increases monotonously. The gamma dose rate given from the radioactive materials deposited on the ground shows the minimum value at 0.07 MeV in any case. The analysis shows that external gamma exposure is strongly dependent on the distribution of radioactive materials in the environment and gamma energy.

• Journal of Radiation Protection and Research
• /
• v.40 no.4
• /
• pp.216-222
• /
• 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.

• Journal of Radiation Protection and Research
• /
• v.19 no.1
• /
• pp.69-80
• /
• 1994

The TCMI(Three-Compartment Model for iodine) computer code has been developed, which is based on the three-compartment model and the respiratory model recommended in ICRP publication 54. This code is able to evaluate the thyroid burden, dose equivalent, committed dose equivalent and urinary excretion rate as time-dependent functions from the input data: working time and the radioiodine concentration in air. Using the TCMI code, the time-dependent thyroid burdens, the thyroid doses and the urinary excretion rates were calculated for three specific exposure patterns : acute, chronic and periodic. Applicability as an internal dose evaluation method has been assessed by comparing the results with some operational experiences. Simple equations and tables are provided to be used in the evaluation of the thyroid burden and the resulting doses for given I-131 concentration in air and the working time.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.1
• /
• pp.53-59
• /
• 2004

A variety of factors such as the pattern of intake (acute or chronic), monitoring interval and the characteristics of the radionuclides could have a significant influence on the estimates for the intake and internal dose. The relative differences of the assessed intakes based on the assumption of an acute intake to that of a chronic intake were evaluated by using the predicted bioassay quantity in the whole body or organs for an acute and chronic intake through the inhalation of $^{125}$ I, $^{137}$ C, $^{235}$ U with the AMAD of 1 ${\mu}{\textrm}{m}$ and 5 ${\mu}{\textrm}{m}$ for the monitoring intervals of 7, 14, 30, 60, 90, 120, 180, 360 days, respectively, The relative difference of the assessed intakes based on the intake pattern is affected by the monitoring interval, radionuclide and absorption type, but the particle size has little influence on the difference of the assessed intakes based on the intake pattern. The maximum monitoring interval, which is defined as the monitoring interval that the relative difference of the assessed intakes based on the assumption of an acute intake to that of a chronic intake is less than 10%, is 60 d for $^{125}$ I with Type F, 180 d for $^{137}$ C with Type F, 90 d for $^{235}$ U with Type M, and 360 d for $^{235}$ U with Type S. It was concluded that an intake pattern has little influence on the estimates of the assessed intake in the case where the monitoring interval is shorter than the maximum monitoring interval for each radionuclide.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.06a
• /
• pp.437-438
• /
• 2004

원자력시설의 정상운영시 방사성물질이 기체 및 액체상의 형태로 소외로 배출되며, 일정기간별로 배출된 핵종별 총량을 이용하여 주민피폭선량 평가를 수행함으로써 규제요건의 준수 여부를 확인한다. 이러한 원자력시설 주변 주민피폭선량평가에 적용되고 있는 주요 입력자료인 음식물섭취자료 등은 1988년 원자력연구소의 현장조사, 실험, 문헌조사 등을 통하여 결정되었으나 시간이 지남에 따라 일부자료의 경우 최신경향을 반영할 수 있도록 개정이 요구된다.(중략)

• Journal of Environmental Impact Assessment
• /
• v.11 no.1
• /
• pp.13-23
• /
• 2002

본 연구에서는 침출수를 비롯한 매립지의 각종 오염물질 배출로 수자원이 오염되어 피폭체의 피해가 빈발하는 문제를 해결하기 위해서, 매립지의 상대적 유해성을 평가하여 한정된 환경관리 예산의 합리적 배분을 위한 우선순위를 결정할 수 있는 의사결정 지원도구로서 LHR(Landfill Site Hazard Ranking)모형을 개발했다. LHR모형은 다요소의사결정(多要素意思決) 기법에 정성적 위해성(危害性) 평가기법을 접맥시켜 주관적 가중치를 모형에 반영한 가치내재화(價値內在化) 모형이다. LHR모형은 피폭체의 주요 피폭경로를 지하수 이동경로와 지표수 이동경로로 보았으며, 각 이동경로별로 누출 가능성, 폐기물 특성 및 피폭체 특성으로 요소범주를 3종류로 구분하여 폐기물의 독성이나 매립량같은 특성이 매립지의 수리지질학적 요소 및 자연지리적 요소에 의해 결정되는 오염물질의 누출 가능성을 통해 매립지 주변의 지역주민과 취약한 수생태계 같은 피폭체에 끼치는 매립지의 유해성을 상대적으로 평가했다. 그리고 LHR모형에서는 매립지 유해성을 공기 이동경로 및 사회경제적 측면에서도 평가하기 위해 매립지 이격거리별 토지이용 형태의 유해성을 평가했다. 그리고 각 평가요소별 가중치는 위계분석과정(位階分析過程)의 쌍대비교법(雙對比較法)에 의하여 할당했으며, 민감도 분석으로 LHR모형을 검증했다.

• Korean Journal of Veterinary Research
• /
• v.41 no.1
• /
• pp.99-104
• /
• 2001

To determine if micronucleus (MN) assay could be used to predict the absorbed dose of victims after accidental radiation exposure, we carried out to assess the absorbed dose depending on the numerical changes of MN in human peripheral blood lymphocytes after $^{60}Co\;{\gamma}-rays$ exposure in the range of 0.25 to 1 Gy, respectively. The MNs were observed at very low doses, and the numerical changes according to doses. Satisfactory dose-effect calibration curve is observed after low dose irradiation of human lymphocytes in vitro. When plotting on a linear scale against radiation dose, the line of best fit was $Y=(0.02{\pm}0.0009)+(0.033{\pm}0.010)D+(0.012{\pm}0.012)D^2$. The dose-response curve for MN induction immediately after irradiation was linear-quadratic and has a significant relationship between the frequencies of MN and dose. These data show a trend towards increase of the numbers of MN with increasing dose. The number of MN in lymphocytes that were observed in the control group is $0.1610{\pm}0.0093/cell$. Accordingly, MN assay in human peripheral lymphocytes could be a useful in viva model for studying radio-protective drug sensitivity or screening test, microdosimertic indicator and radiation-induced target organ injury. Since MN assay is simple, rapid and reproducible, it will also be a biodosimetric indicator for individual dose assessment after accidental exposure.

• Journal of Radiation Protection and Research
• /
• v.32 no.1
• /
• pp.1-8
• /
• 2007

A model for radiological dose assessment in an urban environment, METRO-K has been developed. Characteristics of the model are as follows ; 1) mathematical structures are simple (i.e. simplified input parameters) and easy to understand due to get the results by analytical methods using experimental and empirical data, 2) complex urban environment can easily be made up using only 5 types of basic surfaces, 3) various remediation measures can be applied to different surfaces by evaluating the exposure doses contributing from each contamination surface. Exposure doses contributing from each contamination surface at a particular location of a receptor were evaluated using the data library of kerma values as a function of gamma energy and contamination surface. A kerma data library was prepared fur 7 representative types of Korean urban buildings by extending those data given for 4 representative types of European urban buildings. Initial input data are daily radionuclide concentration in air and precipitation, and fraction of chemical type. Final outputs are absorbed dose rate in air contributing from the basic surfaces as a function of time following a radionuclide deposition, and exposure dose rate contributing from various surfaces constituting the urban environment at a particular location of a receptor. As the result of a contaminative scenario for an apartment built-up area, exposure dose rates show a distinct difference for surrounding environment as well as locations of a receptor.

• Journal of Radiation Protection and Research
• /
• v.28 no.3
• /
• pp.207-213
• /
• 2003

Carbon-14 is one of the major radionuclides released by CANDU Nuclear Power Plants(NPPs). It is almost always emitted as gas through the stack. From CANDU NPPs about 95% of all carbon-14 is released as carbon dioxide. Carbon-14 is a low energy beta emitter which, therefore, gives only a small skin dose from external radiation. As carbon dioxide Is physiologically rather inert gases for man's metabolism, the inhalation dose is probably less than 1 % of the ingestion dose. But this source of carbon-14, formed in a closed, nor-oxidative environment, was subsequently released into the workplace as an insoluble particulate when these systems were opened lip for re-tubing at CANDU NPPs. As a part of the improvement of dosimetry program at Wolsong Nuclear Power Plants, the carbon-14 metabolism based on references was investigated and studied to setup the internal dosimetry program due to inhalation of carbon-14.

• Journal of Radiation Protection and Research
• /
• v.23 no.1
• /
• pp.25-32
• /
• 1998

Tritium is released into the environment as various form but HTO is most relevant to the incorporation of tritium into living organisms and to the formation of OBT. Tritiated compounds may be transported to humans via several tropic levels. Retention times of organically bound tritium in organisms are considerably longer than those of tritiated water, which has significant consequences on exposure dose estimation. It is rather difficult to predict the amount of OBT produced by the photosynthetic process of plants because it may be influenced by a multitude of environmental factors and plant parameters. Tritiated organic matters are classified as the exchangeable or nonexchangeable bound tritium according to the bound form. After short term HTO exposure in plants, the formation and uptake of OBT were evaluated.