• Journal of Radiation Protection and Research
• /
• v.19 no.3
• /
• pp.209-221
• /
• 1994

In the case of a severe accident of a nuclear power plant, the whole body dose and the relative importance of the radionuclides during the lifetime of an exposed person were estimated for each exposure pathway with distances from the release point. The external exposure pathways due to immersion of radioactive cloud and deposition of radioactive materials on the ground, and the internal exposure pathways due to inhalation and ingestion of contaminated foodstuffs were considered. The effects due to the ingestion of contaminated foodstuffs were estimated considering the variation of radioactive concentration in the foodstuffs according to deposition time and elapsed time after deposition using a dynamic ingestion pathway model applicable to Korean environment, named 'KORFOOD'. As the results up to 80 km from the release point, the effects due to ingestion of contaminated foodstuffs showed the highest contribution to total exposure dose. The contribution of I isotopes was the highest in the case of the external dose due to immersion of radioactive cloud and internal dose due to inhalation. The contribution of Cs isotopes was highest in the case of the external dose due to deposition of radioactive materials on the ground. In the case of the internal dose due to ingestion of contaminated foodstuffs, Cs deposition in summer and Sr deposition in winter, respectively, were the most dominant radionuclide to whole body.

• Journal of radiological science and technology
• /
• v.28 no.2
• /
• pp.117-122
• /
• 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.

• Journal of Radiation Protection and Research
• /
• v.36 no.2
• /
• pp.99-106
• /
• 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.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.05b
• /
• pp.913-917
• /
• 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.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.11a
• /
• pp.317-320
• /
• 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.

• 대한방사선방어학회:학술대회논문집
• /
• 2009.11a
• /
• pp.84-85
• /
• 2009

• Journal of Radiation Protection and Research
• /
• v.34 no.3
• /
• pp.129-136
• /
• 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.

• Journal of Radiation Protection and Research
• /
• v.40 no.4
• /
• pp.211-215
• /
• 2015

In case of radiation emergencies, radioactive materials released into environments can cause internal exposure of members of the public. Even though whole body counters are widely used for direct measurement of internally deposited radionuclides, those are not likely to be used at the field to rapidly screen internal exposure. In this study, we estimated the counting efficiencies of portable NaI detector for different size BOMAB phantoms using Monte Carlo transport code to apply handheld gamma spectrometers for rapid screening of internal exposure following radiological accidents. As a result of comparison for two counting geometries, counting efficiencies for sitting model were about 1.1 times higher than those for standing model. We found, however, that differences of counting efficiencies according to different size are higher than those according to counting geometry. Therefore, we concluded that when we assess internal exposure of small size people compared to the reference male, the body size should be considered to estimate more accurate radioactivity in the human body because counting efficiencies of 4-year old BOMAB phantom were about 2.4~3.1 times higher than those of reference male BOMAB phantom.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2012.05a
• /
• pp.369-370
• /
• 2012

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.4
• /
• pp.473-478
• /
• 2018

Radioactive aerosol generated in cutting and melting work during the NPP decommissioning process can cause internal exposure to body through workers' breath. Thus, it is necessary to assess worker internal exposure due to the radioactive aerosol during decommissioning. The actually measured value of the working environment is needed for accurate assessment of internal exposure, but if it is difficult to actually measure that value, the internal exposure dose can be estimated through recommended values such as the fraction of amount of intake and the size of particles suggested by the International Committee on Radiological Protection (ICRP). As for the selection of particle size, this study applied a value of $5{\mu}m$, which is the size of particles considering the worker recommended by the ICRP. As for the amount of generation, the amount of intake was estimated using data on the mass of aerosol generated in a melting facility at a site in Kozloduy, Bulgaria. In addition, using these data, this study calculated the level of radioactivity in the worker's body and stool and conducted an assessment of internal exposure using the BiDAS computer code. The internal exposure dose of Type M was 0.0341 mSv, that of Type S was 0.0909 mSv. The two types of absorption showed levels that were 0.17% and 0.45% of the domestic annual dose limit, respectively.