• Journal of Radiation Protection and Research
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• v.39 no.1
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• pp.61-64
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• 2014

Radioactive materials are in use and have many applications from the generation of electricity to the purposes of research, industry and medicine such as diagnosis and therapy. In the course of their use some of radioactive substances may be discharged into the environment from facilities using the unsealed radioactive materials, which are main artificial sources occurring the public exposure. Discharges are in the form of gases, particles or liquids. This paper provides procedures to estimate the level of the public exposure based on the conservative assumptions and simple calculations in the facility using unsealed liquid sources. They consist of two processes; (1) to calculate maximum concentration of gaseous effluents discharged through the exhaust pipe and average concentration of liquid effluents discharged through the drain of the storage tank, (2) to compare each of them to numerical guidances for the discharges of radioactive gaseous and liquid effluents mentioned in the related notification. For this purpose followings are assumed properly; daily usage, form and dispersion rate of radionuclides, daily amount of radioactive liquid waste and exhaust and drainage equipment. The procedures are readily applicable to evaluate environmental effects by planned effluent discharges from facilities using the unsealed radioactive materials. In addition they may be utilized to obtain practical requirements for radiation safety control necessary for the reductions of the public exposure.

• Journal of Radiation Protection and Research
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• v.6 no.1
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• pp.41-44
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• 1981

A method is presented by which precise evaluation of radioactive concentrations in liquid or gaseous effluent released from large nuclear facilities is possible. Calculations have been made of the calibration constants at different energies for a commonly used liquid and gaseous effluent monitors as well. It is expected that the method could be applicable to the particular monitors with different geometrical configuration with simple modifications.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.1
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• pp.45-54
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• 2024

Currently, off-site dose calculations for nuclear power plants are conducted using a computer program (K-DOSE 60). The program is developed based on the regulatory guidelines of the Korea Institute of Nuclear Safety (KINS), which is a domestic nuclear regulatory agency. In this study, a domestic application of the International Atomic Energy Agency (IAEA) TRS (Technical Reports Series)-472 methodology for ³H and ¹⁴C in liquid effluents was studied. The dose-evaluation methods adopted and the program configuration for dose evaluation are described based on ³H and ¹⁴C in the liquid-effluent-evaluation module of the computer program. The accuracy of the program is verified by comparing the program-calculated results with hand calculation values. Furthermore, a comparative evaluation with LADTAP II, which is a liquid-effluent-evaluation methodology developed by the U.S. NRC (Nuclear Regulatory Commission), is performed. The result confirms that the program-calculated results for the IAEA TRS-472 methodology are consistent with the hand calculation values. Meanwhile, the result of comparative evaluation with LADTAP II indicates different results depending on the methodology used.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1772-1777
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• 2017

Korean nuclear power plants (NPPs) periodically evaluate the radioactive gaseous and liquid effluents released from power reactors to protect the public from radiation exposure. This paper provides a comprehensive overview of the release of radioactive effluents from Korean NPPs and the effects on the annual radiation doses to the public. The amounts of radioactive effluents released to the environment and the resulting radiation doses to members of the public living around NPPs were analyzed for the years 2011-2015 using the Korea Hydro & Nuclear Power Co., Ltd's annual summary reports of the assessment of radiological impact on the environment. The results show that tritium was the primary contributor to the activity in both gaseous and liquid effluents. The averages of effective doses to the public were approximately on the order of $10^{-3}mSv$ or $10^{-2}mSv$. Therefore, even though Korean NPPs discharged some radioactive materials into the environment, all effluents were within the regulatory safety limits and the resulting doses were much less than the dose limits.

• Membrane and Water Treatment
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• v.3 no.4
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• pp.231-242
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• 2012

The performance of a nanofiltration membrane for treatment of a low-level radioactive liquid waste was investigated through static and dynamic tests. The liquid waste ("carbonated water") was obtained during conversion of $UF_6$ to $UO_2$. In the static tests membrane samples were immersed in the waste for 24, 48 or 72 h. The transport properties of the samples (hydraulic permeability, permeate flow, selectivity) were evaluated before and after immersion in the waste. In the dynamic tests the waste was permeated in a permeation flow front system under 0.5 MPa, to determine the selectivity of NF membranes to uranium. The surface layer of the membrane was characterized by zeta potential, field emission microscopy, atomic force spectroscopy and infrared spectroscopy. The static test showed that the pore size distribution of the selective layer was altered, but the membrane surface charge was not significantly changed. 99% of uranium was rejected after the dynamic tests.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.562-571
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• 2003

A series of factors such as sampling, pretreatment measurement, volume estimation which induces uncertainty of the calculated radioactivity in liquid effluent released from a nuclear power plant were analyzed. It is innately impossible to estimate exact error of the calculated radioactivity, since most of the input parameters are determined by a single measurement and true value of the released radioactivity cannot be known. In this paper, a systematic model to calculate uncertainty of the released liquid radioactivity was developed based upon the guidance report published by the ISO in 1993, and the model was applied to a set of hypothetical batch release conditions. As a result, the Priority of each input parameter was turned out to be (1) wastewater volume, (2) sample volume, and (3) measured radioactivity of the sample. In addition, probability distribution of the released radioactivity was simulated by Monte Carlo method combining the probability distribution of each input parameter It was shown that the radioactivity released to the environment, which has been reported as a single value, has a certain form of probability distribution.

• Journal of Radiation Protection and Research
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• v.39 no.1
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• pp.38-45
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• 2014

This paper describes the results of assessment of radiological dose resulting from operation of the Daedeok nuclear facilities including the HANARO research reactor, which has been performed to assure whether or not to comply with the regulation standards of the radioactive effluents releases. Based on the meteorological data and the radiation source term, the maximum individual doses were evaluated from 2010 to 2012. The atmospheric dispersion and the deposition factors of gaseous effluents were calculated using the XOQDOQ computer code. ENDOS-G and ENDOS-L code systems were also used for maximum individual dose calculation from gaseous and liquid effluents, respectively. The results were compared with the regulation standards for the radioactive effluents presented by the Nuclear Safety and Security Commission (NSSC). The effective doses and the thyroid doses of the maximum individual were calculated at the maximum exposed point in the Daedeok site, and contributions of exposure pathways to the radiological doses resulting from gaseous and liquid radioactive effluents were evaluated at each facility of the Daedeok site. As a result, the maximum exposed age was analysed to be the child group, and the operation of HANARO research reactor had a major effect more than 90% on the individual doses. The main exposure pathways for gaseous radioactive effluent were from ingestion and inhalation. The effective doses and the thyroid doses were considerably influenced by tritium and iodine, respectively. The gaseous radioactive effluents contributed more than 90% on the total doses, whereas the contributions of the liquid radioactive effluents were relatively low. Consequently, the maximum individual dose due to radioactive effluents from the nuclear facilities within the Daedeok site were less than 3% of the regulation standard over 3 years; therefore, it can be concluded that radioactive effluents from the nuclear facilities were well managed, with the radiation-induced health detriment for residents around the site being negligible.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.321-331
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• 2005

Recently. three-dimensional models have been used for aquatic dispersion of radioactive effluents in relation to nuclear power plant siting based on the Notice No. 2003-12 'Guideline for investigating and assessing hydrological and aquatic characteristics of nuclear facility site' of the Ministry of Science and Technology (MOST) in Korea. Several nuclear power plants have been under construction or planed. which are Shin-Korl Unit 1 and 2, Shin-Wolsong Unit 1 and 2, and Shln-Ulchin Unit 1 and 2. For assessing the aquatic dispersion of radionuclides released from the above nuclear power plants, it is necessary to know the coastal currents around sites which are affected by circulation of East Sea. In this study, a three dimensional hydrodynamic model for the circulation of the East Sea of Korea has been developed as the first Phase, which Is based on the RIAMOM. The model uses the primitive equation with hydrostatic approximation, and uses Arakawa-B grid system horizontally and Z-coordinate vertically. Model domain is $126.5^{\circ}E\;to\;142.5^{\circ}E$ of east longitude and $33^{\circ}N\;and\;52^{\circ}N$ of the north latitude. The space of the horizontal grid was $1/12^{\circ}$ to longitude and latitude direction and vortical level was divided to 20. This model uses Generalized Arakawa Scheme. Slant Advection, and Mode-Splitting Method. The input data were from JODC, KNFRDI, and ECMWF. The model ing results are in fairly good agreement with schematic patterns of the surface circulation in the East Sea The local current model and aquatic dispersion model of the coastal region will be developed as the second phase. The oceanic dispersion experiments will be also tarried out by using ARGO Drifter around a nuclear pelter plant site.

• Nuclear Engineering and Technology
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• v.6 no.2
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• pp.97-111
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• 1974

The adsorption mechanism of Cs-137 in low level radioactive solution by vermiculite treated with Na ion is studied in order to investigate its effective utilization for the radioactive effluent treatment. The beneficial role of Na-vermiculite is that Na ion can induce the wider c-axis spacing in which Cs ion can be sorbed in vermiculite. Cation exchange capacity and distribution coefficient of cesium seems to be influenced by the variation of c-axis spacing of vermiculite. Comparative identification and detection with the characteristic analyses of X-ray diffraction and electron diffraction patterns, diffrential thermal analysis and electron microscopy of Na-, K- and Cs-vermiculite are studied for the phemomena of Cs adsorption by vermiculite. This importance of the utilization in terms of adsorption and fixation of cesium involving vermiculite is discussed. It is found that the Na-vermiculite is valuable outside charging material for high level radioactive liquid waste storage tank of underground to protect the pollution of the underground water.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.1-10
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• 2015

After the Fukushima Daiichi nuclear power plant (NPP) accident, new regulatory requirements were enforced in July 2013 and a backfit was required for all existing nuclear power plants. It is required to take measures to prevent severe accidents and mitigate their radiological consequences. The Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R) has been conducting numerical studies and experimental studies on relevant severe accident phenomena and countermeasures. This article highlights fission product (FP) release and hydrogen risk as two major areas. Relevant activities in the S/NRA/R are briefly introduced, as follows: 1. For FP release: Identifying the source terms and leak mechanisms is a key issue from the viewpoint of understanding the progression of accident phenomena and planning effective countermeasures that take into account vulnerabilities of containment under severe accident conditions. To resolve these issues, the activities focus on wet well venting, pool scrubbing, iodine chemistry (in-vessel and ex-vessel), containment failure mode, and treatment of radioactive liquid effluent. 2. For hydrogen risk: because of three incidents of hydrogen explosion in reactor buildings, a comprehensive reinforcement of the hydrogen risk management has been a high priority topic. Therefore, the activities in evaluation methods focus on hydrogen generation, hydrogen distribution, and hydrogen combustion.