• Journal of Radiation Protection and Research
• /
• v.28 no.3
• /
• pp.247-254
• /
• 2003

To prevent the potential health detriment to the public from radioactive effluents, radiological dose assessments due to the operation of nuclear facilities located at Korea Atomic Energy Research Institute (KAERI) site has been performed semiannually in compliance with the Minister of Science and Technology (MOST)'s Notice in Korea. Radiological dose assessment based on the new recommendation of the International Committee on Radiation Protection (ICRP-60) has been conducted since 1998. In this manuscript, a serial activities at KAERI site to meet the regulatory standards for routine releases of radioactive effluents are introduced and discussed including technical approaches. It is clear that each nuclear facility has been operated in compliance with regulatory standards. Furthermore, it is identified that the radiation induced health effects for residents around the site are neglectable.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.2
• /
• pp.247-253
• /
• 2023

According to NSSC Notice No. 2021-10, safety analysis needs to be introduced in the decommissioning plan. Public and occupational dose analyses should be conducted, specifically for unexpected radiological accidents. Herein, based on the risk matrix and analytic hierarchy process, the method of selecting accident scenarios during the decommissioning of nuclear power plants has been proposed. During decommissioning, the generated spent resin exhibits relatively higher activity than other generated wastes. When accidents occur, the release fraction varies depending on the conditioning method of radioactive waste and type of radioactive nuclides or accidents. Occupational dose analyses for 2 (fire and drop) among 11 accident scenarios have been performed. The radiation doses of the additional exposures caused by the fire and drop accidents are 1.67 and 4.77 mSv, respectively.

• The Korean Journal of Zoology
• /
• v.3 no.1
• /
• pp.19-23
• /
• 1960

A method for determination of thyroid secretion rate in rabbit by means of radioactive iodine presented. After injection of radioactive iodine, in vivo determination so f radioactivity in thyroid gland were made during a 19 day-experimental period. In the same period blood samples were drawn and analyzed for protein-bound iodine (PBI) and for protein-bound radioactive iodine(PBI181). A rate constant for secretion of thyroid hormone was calculated from the disappearance rate of radioactive iodine in thyroid gland. The secretion rate of radioactive hormone iodine was calculated by multiplying this rate constant by the amount of radioactive iodine present in thyroid gland. Assuming that the specific radioactiveness of the circulating thyroid hormone and of the hormone just secreted were identical , thyroid secretion rate was calculated by the equation. {{{{ { Secreted hormone-iodine , gamma /hr} over { Secreted hormone-I^131, % dose/hr }= { PBI, ${\gamma}$/ml.Serum} over { PBI^131 , % dose/ml . Serum } }} The method presented consisted of measurements for series of independent criteria on thyroid function, and the resulting thyroid secretion rate was calculated by combination of those.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.18 no.2
• /
• pp.157-167
• /
• 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.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.579-583
• /
• 2003

This study was to compare the Intake and Committed Effective Dose(CED) due to inhalation of I-131. The Intake and CED Calculations were based on KIDAC, LUDEP, MONDAL and IMBA computer codes.

• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.293-300
• /
• 2022

During the decommissioning of the nuclear facilities, the radioactive gases in pressure vessels may leak due to the demolition operations. The decommissioning site has large space, slow air circulation, and many large nuclear facilities, which increase the difficulty of workers' inhalation exposure assessment. In order to dynamically evaluate the activity distribution of radionuclides and the committed effective dose from inhalation in nuclear decommissioning environment, an inhalation exposure assessment method based on the modified eddy-diffusion model and the inhaled dose conversion factor is proposed in this paper. The method takes into account the influence of building, facilities, exhaust ducts, etc. on the distribution of radioactive gases, and can evaluate the influence of radioactive gases diffusion on workers during the decommissioning of nuclear facilities.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.2
• /
• pp.255-269
• /
• 2023

Compared to operational wastes, nuclear power plant (NPP) decommissioning wastes are generated in larger quantities within a short time and include diverse types with a wider range of radiation characteristics. Currently used 200 L drums and IP-2 type transport containers are inefficient and restrictive in packaging and transporting decommissioning wastes. Therefore, new packaging and transport containers with greater size, loading weight, and shielding performance have been developed. When transporting radioactive materials, radiological safety should be assessed by reflecting parameters such as the type and quantity of the package, transport route, and transport environment. Thus far, safety evaluations of radioactive waste transport have mainly targeted operational wastes, that have less radioactivity and a smaller amount per transport than decommissioning wastes. Therefore, in this study, the possible radiation effects during the transport from NPP to disposal facilities were evaluated to reflect the characteristics of the newly developed containers and decommissioning wastes. According to the evaluation results, the exposure dose to transport workers, handling workers, and the public was lower than the domestic regulatory limit. In addition, all exposure dose results were confirmed, through sensitivity analysis, to satisfy the evaluation criteria even under circumstances when radioactive materials were released 100% from the container.

• The Korean Journal of Nuclear Medicine
• /
• v.9 no.1
• /
• pp.59-71
• /
• 1975

53 patients with hyperthyroidism have been analyzed with special reference to therapeutic response to radioactive iodine ($^{131}I$) treatment. Mean effective half-life, 24 hour uptake rate and radiation dose of $^{131}I$ in hyperthyroid patients included in this study were respectively. 1. Mean effective half-life of $^{131}I\;was\;4.7{\pm}1.5$ days in the tracer dose and $5.0{\pm}1.5$ days in the therapeutic dose. 2. Mean 24 hour uptake rate of $^{131}I\;was\;72.7{\pm}11.1%$ in the tracer dose and $73.4{\pm}12.3%$ in the theapeutic dose. 3. Mean radiation dose of $^{131}I\;was\;5,319{\pm}2,648$ RAD as predicted and $5,692{\pm}2,843$ RAD as actual. A single dose of radioactive iodine treatment was satisfactory in 34 patients (radioiodine sensitive) and multiple doses of radioactive iodine treatments were required in 19 patients (radioiodine resistant). A radioiodine resistant group of patients with hyperthyroidism was distinctively characteristic in the following aspects. 1. Mean thyroid weight calculated in the resistant group ($63.9{\pm}14.0gm$) was significantly (p<0.01) greater than that of the sensitive group ($46.6{\pm}13.3gm$). 2. Mean 24 hour uptake rate of the tracer dose in the resistant group ($67.3{\pm}10.7%$) was significantly (p<0.01) lower than that of the sensitive group ($75.7{\pm}10.5%$). 3. Mean 24 hour uptake rate of the therapeutic dose in the resistant group ($68.5{\pm}13.7%$) was significantly (p<0.05) lower than that of the sensitive group ($76.1{\pm}10.9%$). 4. Mean predicted radiation dose, of $^{131}I$ in the resistant group ($3,684{\pm}1,745$ RAD) was significantly (p<0.01) lower than that of the sensitive group ($6,232{\pm}2,683$ RAD). 5. Mean actual radiation dose of $^{131}I$ in the resistant group ($4,100{\pm}1,691$ RAD) was significantly (p<0.01) lower than that of the sensitive group ($6,582{\pm}3,024$ RAD). 6. No significant difference was detected in terms of effective half-life of $^{131}I$ among the groups (p>0.05). 7. The average mean % difference of effective half-life, uptake rate and radiation dose measured following the tracer and therapeutic dose of $^{131}I$ were not statistically significant (p>0.05). Therefore effective half-life, uptake rate and radiation dose of the therapeutic dose of $^{131}I$ were readily predictable following the tracer dose of $^{131}I$. 8. It is concluded that the possibility of resistance to radioactive iodine treatment may be anticipated in patients with thyroid gland large in size and compromised $^{131}I$ uptake rate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.1
• /
• pp.79-90
• /
• 2016

The second-stage near surface disposal facility for low and very low level radioactive waste's permanent disposal is to be built. During the institutional control period, the inadvertent intrusion of the general public is limited. But after the institutional control period, the access to the general public is not restricted. Therefore human who has purpose of residence and resource exploration can intrude the disposal facility. In this case, radioactive effects to the intruder should be limited within regulatory dose limits. This study conducted the safety assessment of human intrusion on the second-stage surface disposal facility through drilling and post drilling scenario. Results of drilling and post drilling scenario were satisfied with regulatory dose limits. The result showed that post-drilling scenario was more significant than drilling scenario. According to the human intrusion time and behavior after the closure of the facility, dominant radionuclide contributing to the intruder was different. Sensitivity analyses on the parameters about the human behavior were also satisfied with regulatory dose limits. Especially, manual redistribution factor was the most sensitive parameter on exposure dose. A loading plan of spent filter waste and dry active waste was more effective than a loading plan of spent filter waste and other wastes for the radiological point of view. These results can be expected to provide both robustness and defense in depth for the development of safety case further.

• Journal of the Korea Institute of Information and Communication Engineering
• /
• v.24 no.11
• /
• pp.1519-1527
• /
• 2020

AGeneral radiation measuring devices have been developed in the form of spatial dose rate detection devices that measure dose rates to radioactive contaminant and 2D or 3D imaging devices for radioactive contamination information. Each of these radiation detection techniques has advantages. The advantages of both detection devices are necessary to minimize personal injury and rapid decontamination in the area of a radioactive accident. In this paper, we proposed a technique that can measure the dose rate and direction information about the radioactive pollutant source in real time using a detection sensor, a rotating body, and a directional shield for radioactive pollutant detection. The rotational-based detection device is configured to check the dose rate and direction using the location information of the rotator and measurement value. We proposed a measurement technique for vertical and horizontal directions through multiple holes. It was confirmed that the measurement error for direction information was less than 1% when detected in the horizontal direction.