• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.201-209
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• 2016

The object of this study was to evaluate the separation distance from a temporary storage facility satisfying the dose criteria. The calculation of ambient dose rates took into account cover soil thickness, facility size, and facility type by using MCNPX code. Shielding effects of cover soil were 68.9%, 96.9% and 99.7% at 10 cm, 30 cm and 50 cm respectively. The on-ground type of storage facility had the highest ambient dose rate, followed by the semi-ground type and the underground type. The ambient dose rate did not vary with facility size (except $5{\times}5{\times}2m\;size$) due to the self-shielding of decontamination waste in temporary storage. The separation distances without cover soil for a $50{\times}50{\times}2m\;size$ facility were evaluated as 14 m (minimum radioactivity concentration), 33 m (most probably radioactivity concentration), and 57 m (maximum radioactivity concentration) for on-ground storage type, 9 m, 24 m, and 45 m for semi-underground storage type, and 6 m, 16 m, and 31 m for underground storage type.

• 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.

• Progress in Medical Physics
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• v.18 no.4
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• pp.187-193
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• 2007

In an effort to assess the internal absorbed dose of radionuclides that is suitable to Koreans' physiological characteristics, we asked 28 male Koreans to take $^{131}|$ orally, determined the thyroidal uptake and daily urination ratio, and assessed the absorbed dose by organ. As a result, first, 24 hours after administering, the average thyroidal uptake and the daily urination ratio registered 19.70% and 71.12%, respectively. Second, the whole body effective dose according to the thyroidal uptake calculated herein and the existing ICRP-suggested thyroidal uptake of 30% offered 1.464E-08 Sv and 2.189E-08 Sv, respectively, showing a 1.5 times difference. To evaluate the quantity of the absorbed dose of radioactive iodine, we can better reduce the error in assessing the body exposure dose by conducting measurement according to human races rather than depending on the existing ICRP data.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.234-235
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• 2011

• Journal of Radiation Protection and Research
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• v.34 no.4
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• pp.176-183
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• 2009

Maintenance on the water chamber of steam generator during outage in nuclear power plants (NPPs) has a likelihood of high radiation exposure to whole body of workers even short time period due to the high radiation exposure rates. In particular, it is expected that hands would receive the highest radiation exposure because of its contact with radiation materials. In this study, characteristic analysis of inhomogeneous radiation fields for contact operations was conducted using thermoluminescent dosimeter (TLD) readouts from the application tests of two-dosimeter algorithm to Korean NPPs in 2004. It is regarded that inhomogeneous radiation fields for contact operations in NPPs are dominated by high energy photons. In addition, field tests for workers who participated in maintenance on the steam generator during outage at Ulchin NPPs in 2009 and pressure tube replacement at Wolsong NPPs in 2009 were conducted to analyze radiation fields and to estimate the extremity dose. As a result, radiation fields were dominated by high energy photons.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.60-67
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• 2004

The exposure dose form recycling of a large amount of the steel scrap from the KRR-1＆2 decommissioning activities was evaluated, and also the clearance level(draft) was derived. The maximum individual dose and collective dose were evaluated by modifying internal dose conversion factor which was based on the concept of effective dose in ICRP 60, applied to the RESRAD-RECYCLE ver 3.06 computing code, IAEA Safety Series 111-P-1.1 and NUREG-1640 as the assessment tool. The result of assessment for individual dose and collective dose is 23.9 $\mu$Sv per year and 0.11 man$.$Sv per year respectively. The clearance levels were ultimately determined by extracting the most conservative value form the results of the generic assessment and specific assessment methodologies. The result of clearance level for radionuclides( $Co^{60}$ , C $s^{l37}$) is less than 1.14${\times}$10$^{-1}$ Bq/g to comply with the clearance criterion(maximum individual dose : 10 $\mu$Sv per year, collective dose : 1 man$.$Sv per year) provided for Korea Atomic Energy Act and relevant regulations.s.

• Nuclear Engineering and Technology
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• v.7 no.3
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• pp.191-198
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• 1975

A study was made on the neutron dosimetry in a mixed gamma-neutron field with LiF thermoluminescent dosimeter. In order to estimate the neutron dose in a mixed field, $^{6}$ LiF and $^{7}$ LiF dosimeters were used for fast and thermal neutron doses. The over-all conversion factors for the effects of dosimeter positions were derived for personnel monitoring and the glow curves of the LiF dosimeters for neutron and gamma-ray doses were also analyzed.

• Journal of Radiation Protection and Research
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• v.11 no.1
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• pp.37-43
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• 1986

Dosimetric quantities for 300 keV neutrons in the ICRU standard tissue sphere were evaluated. The Monte Carlo code NEDEP which performs neutron-photon-charged particles coupled transport was used in the direct estimation of absorbed dose and dose equivalent. Some important quantities calculated are as follows; Deep dose equivalent index $H_{I,d}:1.78{\times}10^{11}\;Sv-cm^2$ Shallow dose equivalent index $H_{I,s}:2.08{\times}10^{-11}\;Sv-cm^2$ Ambient dose equivalent $H^*(0.07):1.7{\times}10^{-11}\;Sv-cm^2$ Ambient dose equivalent $H^*(10):1.78{\times}10^{-11}\;Sv-cm^2$ Effective quality factor $\bar{Q}^*(10):12.4$

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.497-502
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• 2003

Landfill is an widely applied alternative for clearance of non-contaminated waste or slightly contaminated waste generated from nuclear facilities. In this study, exposure dose is estimated for a worker and a resident at the landfill area in Young-gwang nuclear power plant. Based on evaluated dose, clearance concentrations of each radionuclide are determinated for dose criteria of 10 $\muSv/y$. The results of age-dependent dose are 1.02 $\muSv$ per year for resident and 0.471 $\muSv$ per year for worker. Clearance concentrations for each radionuclide are evaluated from $1.33{\times}10_{-1}$ Bq per gram to $2.85{\times}10^2$ Bq per gram.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.22-22
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• 2005