• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.234-236
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• 2013

Some of radioactive contamination is unavoidable in the facilities using the unsealed radioactive material. The primary purpose of radioactive contamination control in the workplace with contamination concern is the effects from the potential intake of radioactive material into the body. This paper provides procedures to estimate the level of internal exposure for the worker based on the conservative assumptions and simple calculations. They consist of two processes; to calculate air concentration of radioactive material and annual intake by inhalation with contaminated air and to compare each of them to Derived Air Concentration and Annual Limit on Intake mentioned in the related notification. The procedures are applicable to make a decision on practical requirements for monitoring air contamination and internal exposure of worker as follows; needs for measurement of air contamination and internal exposure and acquisition of information on the design of the ventilation system.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.267-278
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• 1985

Some kinds of methods have been applied to regulate the exposure doses by the radioactive effluents from nuclear power plants. The essential one is primary dose equivalent limit recommended by the ICRP. When the primary limit cannot be applied directly for regulation, there have been dose equivalent index in case of external exposure, or maximum permissible concentration, annual limit on intake, derived air concentration and maximum permissible body burden in case of internal exposure. But the derived limit is required from the viewpoint of discharge, for those values are inadequate to control discharge rate directly. This study was carried out to derive the release limit for the Wolsung nuclear power plant by the concentration factor method. This method is based on the assumption of steady state transfer between environment compartments.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2050-2056
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• 2024

Handling of radioisotopes may cause external and internal contamination to occupational workers while using radiation for medical purposes. This research aims to monitor the internal hazard of occupational workers who handle ¹³¹I. Two methods are used: in vivo or direct method and in vitro or indirect method. The in vivo or direct method was performed by assessing thyroid intake with a thyroid uptake monitoring machine. The in vitro or indirect method was performed by assessing urine samples with the help of a gamma-ray spectroscopy practice using a High-Purity Germanium (HPGe) Detector. In this study, fifty-nine thyroid counts and fifty-nine urine samples were collected from seven occupational workers who were in charge of ¹³¹I at the National Institute of Nuclear Medicine and Allied Sciences (NINMAS), Dhaka. The result showed that the average annual effective dose of seven workforces from thyroid counts were 0.0208 mSv/y, 0.0180 mSv/y, 0.0135 mSv/y, 0.0169 m Sv/y, 0.0072 mSv/y, 0.0181 mSv/y, 0.0164 mSv/y and in urine samples 0.0832 mSv/y, 0.0770 mSv/y, 0.0732 mSv/y, 0.0693 mSv/y, 0.0715 mSv/y, 0.0662 mSv/y, 0.0708 mSv/y.The total annual effective dose (in vivo and in vitro method) was found among seven workers in average 0.1039 mSv/y, 0.0950 mSv/y, 0.0868 mSv/y, 0.0862 mSv/y, 0.0787 mSv/y, 0.0843 mSv/y, 0.0872 mSv/y. Following the rules of the International Commission on Radiological Protection (ICRP), the annual limit of effective dose for occupational exposure is 20 mSv per year and the finding values from this research work are lesser than this safety boundary.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.299-303
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• 2001

Using ultrasound techniques, the Korea Atomic Energy Research Institute has measured chest wall thicknesses of a group of male workers at the Korea Atomic Energy Research Institute. A site-specific biometric equation has been developed for these workers. Chest wall thickness is an important modifier on lung counting efficiency. These data have been put into the perspective of the ICRP recommended dose limits for occupationally exposed workers: 100 mSv in a 5-year period with a maximum of 50 mSv in anyone year. For measured chest wall thicknesses of 1.9 cm to 4.1 cm and a 30 min counting time, the achievable MDAs for natural uranium in the KAERI lung counter vary from 5.75 mg to 11.28 mg. These values are close to, or even exceed, the predicted amounts of natural uranium that will remain in the lung (absorption type M and S) after an intake equal to the Annual Limit on Intake corresponding to a committed dose of 20 mSv. This paper shows that the KAERI lung counter probably cannot detect an intake of Type S natural uranium in a worker with a chest wall thickness equal to the average value (2.7 cm) under routine counting conditions.

• Journal of The Korean Society of Agricultural Engineers
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• v.53 no.1
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• pp.17-28
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• 2011

This study was performed to ascertain the possibility of securing inflows to reservoir with low ratio of watershed to paddy field areas by outside diversion weir. The case of Maengdong reservoir and Samryong diversion weir was selected. Most of inflows to Maengdong reservoir with watershed area of $7.06\;km^2$ and total storage capacity of $1,269{\times}10^4\;m^3$ are filled with intake water from outside Samryong diversion weir. Only using water storage data in Maengdong reservoir from 1991 to 2009, the range of water intake in Samryong diversion weir to Maengdong reservoir was optimized to 0.135~30 mm/d, from which water intake to Maengdong reservoir was $1,672.9{\times}10^4\;m^3$ (70.1 %) and downstream outflow to Weonnam reservoir was $714.4{\times}10^4\;m^3$ (29.9 %). The parameters of DAWAST model for reservoir inflow were determined to UMAX of 313.8 mm, LMAX 20.3 mm, FC 136.8 mm, CP 0.018, and CE 0.007. Inflows to Maengdong reservoir were $427.1{\times}10^4\;m^3$ (20.3 %) from inside watershed, and $1,672.9{\times}10^4\;m^3$ (79.7 %) from outside. Paddy irrigation water requirements were estimated to $1,549{\times}10^4\;m^3$ on annual average. Operation rule curve was drawn by using daily inflow and irrigation requirement data. By securing the amount of inflow to Maengdong reservoir to about 80 % from outside Samryong diversion weir, water supply capacity for irrigation of $1,549{\times}10^4\;m^3/yr$ was analyzed to be enough. Additional water supplies for instream flow were analyzed to $1,412\;m^3/d$ in normal reservoir operation, $36,000\;m^3/d$ in withdrawal limit operation by operation rule curve from October to March of non irrigation period.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.291-300
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• 2016

Background: It is necessary to assess radiation dose to workers due to inhalation of airborne particulates containing naturally occurring radioactive materials (NORM) to ensure radiological safety required by the Natural Radiation Safety Management Act. The objective of this study is to develop an internal dose assessment procedure for workers at industries using raw materials containing natural radionuclides. Materials and Methods: The dose assessment procedure was developed based on harmonization, accuracy, and proportionality. The procedure includes determination of dose assessment necessity, preliminary dose estimation, airborne particulate sampling and characterization, and detailed assessment of radiation dose. Results and Discussion: The developed dose assessment procedure is as follows. Radioactivity concentration criteria to determine dose assessment necessity are $10Bq{\cdot}g^{-1}$ for $^{40}K$ and $1Bq{\cdot}g^{-1}$ for the other natural radionuclides. The preliminary dose estimation is performed using annual limit on intake (ALI). The estimated doses are classified into 3 groups ( < 0.1 mSv, 0.1-0.3 mSv, and > 0.3 mSv). Air sampling methods are determined based on the dose estimates. Detailed dose assessment is performed using air sampling and particulate characterization. The final dose results are classified into 4 different levels ( < 0.1 mSv, 0.1-0.3 mSv, 0.3-1 mSv, and > 1 mSv). Proper radiation protection measures are suggested according to the dose level. The developed dose assessment procedure was applied for NORM industries in Korea, including coal combustion, phosphate processing, and monazite handing facilities. Conclusion: The developed procedure provides consistent dose assessment results and contributes to the establishment of optimization of radiological protection in NORM industries.

• Journal of Radiation Industry
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• v.8 no.2
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• pp.123-132
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• 2014

After the nuclear accident of the Fukushima Dai-ichi Nuclear Power Plants (FDNPPs) on 11 March 2011, a large amount of radioactive materials has been released into the atmosphere and the ocean. A compartment model is used to evaluate the circulation characteristics and the spatiotemporal concentration distributions of radionuclides in the ocean. In the comparison with observed concentrations of $^{137}Cs$ in seawater, calculated concentrations by the compartment model were well agreed with them. On the basis of these results, we performed evaluation of the effective dose and the cancer risk. In the early stage of the accident, the effective doses from ingestion of the seafood near the Fukushima region were much higher than 1 mSv which is the value of the annual effective dose limit to individual recommended by the International Commission on Radiological Protection (ICRP). However, the effective doses by ingestion of the seafood decreased below 1 mSv as distance from the FDNPPs increased and time passed. In addition, it was estimated that the cancer risks by intake of the contaminated marine products were less than natural occurrence probability of cancer. Consequently, it was inferred that the health risk due to the $^{137}Cs$ was low after since mid-term period of the accident.

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.747-752
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• 2007

The Proton Engineering Frontier Project (PEFP) has designed and developed a proton linear accelerator facility operating at 100 MeV - 20 mA. The radiological effects of such a nuclear facility on the environment are important in terms of radiation safety. This study estimated the production rates of radionuclides in the soil around the accelerator facility using MCNPX. The groundwater migration of the radioisotopes was also calculated using the Concentration Model. Several spallation reactions have occurred due to leaked neutrons, leading to the release of various radionuclides into the soil. The total activity of the induced radionuclides is approximately $2.98{\times}10^{-4}Bq/cm^3$ at the point of saturation. $^{45}Ca$ had the highest production rate with a specific activity of $1.78{\times}10^{-4}Bq/cm^3$ over the course of one year. $^3H$ and $^{22}Na$ are usually considered the most important radioisotopes at nuclear facilities. However, only a small amount of tritium was produced around this facility, as the energy of most neutrons is below the threshold of the predominant reactions for producing tritium: $^{16}O(n,\;X)^3H$ and $^{28}Si(n,X)^3H$ (approximately 20 MeV). The dose level of drinking water from $^{22}Na$ was $1.48{\times}10^{-5}$ pCi/ml/yr, which was less than the annual intake limit in the regulations.

• Korean Journal of Fisheries and Aquatic Sciences
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• v.11 no.3
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• pp.129-138
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• 1978

The contents of inorganic conservative constituents in the downstream water were determined in spring tides of every month from May 1977 to April 1978 at eight stations of Nagdong River. Samples were taken at the intervals of one or two hours from 7 a. m. to 7 p. m. at each station. Annual ranges and means of the chemical constituents over all the stations except station one, Kupo, ate as follows: pH 6.4-9.3, 7.7; electrical conductivity $0.085-0.345\times10^3\;\mu\mho/cm,\;0.196\times10^3\;\mu\mho/cm;$ chloride 5.8-50.0ppm, 17.7ppm: fluoride ND-0.19 ppm, 0.06 ppm: sulfate 5.5-41.1 ppm, 20.7 ppm; calcium 6-26 ppm, 17 ppm; magnesium 2.0-12.8 ppm, 5.1 ppm; sodium 7-26 ppm, 13 ppm; potassium 1.4-3.8 pprn, 2.3 ppm respectively. The seasonal variations of contents of the chemical constituents were not large and showed nearly definite values at all the stations except station one, Kupo. At station one, seasonal variations were large and the contents were excessively high due to inflow of seawater compared with other stations. The values over 50 ppm in chloride were not determined during the determination period at Mul Geum where the intake station for Busan city water is located. Most constituents except pH and fluoride were over the criteria for drinking water at Kupo, while at other stations only pH value was exceeded the upper limit of the criterion especially in summer period. The pH values tended to increase in the afternoon when water temperature was high. The chloride concentration was shown the highest value at station one, Kupo, with about 2 hours delay after high water of Busan harbour and 3-3.5 hours at Mul Geum.

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