• Journal of Soil and Groundwater Environment
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• v.23 no.4
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• pp.1-15
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• 2018

Despite of safety issues related to radiological hazards, 31 countries around the world are operating more than 450 nuclear power plants (NPPs). To operate NPPs safely, safety regulations from radiation protection organizations were developed and adopted in many countries. However, many cases of radionuclide releases at foreign NPPs have been reported. Almost all commercial NPPs routinely release radioactive materials to the surrounding environments as liquid and gas phases under control. These releases are called 'planned releases' which are planned, regularly monitored, and well documented. Meanwhile, the releases focused in this review, called 'unplanned releases', are neither planned nor monitored by regulatory and/or protection organizations. NPPs are generally composed of various structures, systems and components (SSCs) for safety. Among them, the SSCs near reactors are closely related to safety of NPPs, and typically fabricated to comply with stringent requirements. However, some non-safety related SSCs such as underground pipes may be constructed only according to commercial standards, causing the leakage of radioactive fluids usually containing tritium ($^3H$). This paper discusses SSCs of NPPs and introduces several cases of unplanned releases at foreign NPPs. The current regulation on the environmental radiological surveillance and assessment around the NPPs in South Korea are also examined.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.280.2-280
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• 2002

• Resources Recycling
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• v.31 no.5
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• pp.20-25
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• 2022

In this study, the applicability of microalgae was evaluated for eco-friendly decontamination of cesium-137 (Cs-137) and strontium-90 (Sr-90), which are radioactive nuclides contained in radioactive waste. The monolithic radioactive solution used in the experiment was manufactured at a concentration of 1.5 Bq/mL Cs-137 and 1.0 Bq/mL Sr-90 by diluting a standard radioactive solution and distilled water. This experiment used two types of microalgae, Chlorella Vulgaris was used for Sr-90 decontamination and Hematococcus pluvialis for Cs-137 decontamination. The experimental method is to put the microalgae cultured for 2 weeks into a bottle with a semi-permeable membrane, and then put the bottle in which the microalgae was put into the manufactured radioactive solution, so that the microalgae and the radioactive solution react through the semi-permeable membrane for 48 hours. For the radioactivity concentration analysis of each sample, a gamma-ray nuclide analyzer was used for Cs-137, a γ-ray isotope, and a Liquid Scintillation Count(LSC) was used f or Sr-90, a β-ray isotope. As a result of the experiment, it was confirmed that about 88.0 % of Cs-137 and about 89.7 % of Sr-90 could be decontaminated, and about 98.6 % of Sr-90 was finally able to be decontaminated by the two-stage decontamination method.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.105-108
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• 2003

The intervention area located at rear hot cell can be contaminated by hot cell maintenance work. For effective decontamination of the intervention floor a wet type decontamination device was developed. The device was assembled with a brush rotating part, a washing liquid supplying part, an intake part for recovering contaminated liquid and a device moving cart part. The device was made of stainless steel for easy decontamination and corrosion resistance. The function test carried out at intervention area of the PIE facility showed good performance.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.1-14
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• 2017

For the removal of cesium (Cs) from high radioactive/high salt-laden liquid waste, this study synthesized a highly efficient composite adsorbent (potassium cobalt ferrocyanide (PCFC)-loaded chabazite (CHA)) and evaluated its applicability. The composite adsorbent used CHA, which could accommodate Cs as well as other molecules, as a supporting material and was synthesized by immobilizing the PCFC in the pores of CHA through stepwise impregnation/precipitation with $CoCl_2$ and $K_4Fe(CN)_6$ solutions. When CHA, with average particle size of more than $10{\mu}m$, is used in synthesizing the composite adsorbent, the PCFC particles were immobilized in a stable form. Also, the physical stability of the composite adsorbent was improved by optimizing the washing methodology to increase the purity of the composite adsorbent during the synthesis. The composite adsorbent obtained from the optimal synthesis showed a high adsorption rate of Cs in both fresh water (salt-free condition) and seawater (high-salt condition), and had a relatively high value of distribution coefficient (larger than $10^4mL{\cdot}g^{-1}$) regardless of the salt concentration. Therefore, the composite adsorbent synthesized in this study is an optimized material considering both the high selectivity of PCFC on Cs and the physical stability of CHA. It is proved that this composite adsorbent can remove rapidly Cs contained in high radioactive/high salt-laden liquid waste with high efficiency.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.433-433
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• 2004

Neutron induced prompt gamma-ray spectroscopy (NIPS) system measures the prompt gamma-ray, emitting by the interaction of a neutron with various materials. This system will be of great benefit to scientists worldwide, since it provides the non-destructive measurement of many elements in either solid or liquid wastes. A NIPS facility has been developed in Nuclear Chemistry Research Division, at Korea Atomic Energy Research Institute (KAERI) with the aim of analyzing the major component elements in both aqueous and solid samples.(omitted)

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.15-20
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• 2003

During operation process of SIES(Selective ion exchange system) at Kori Unit 2, it was impossible to remove radionuclides such as ion form and Ag-110m, etc., because activated carbon and ion exchange resin of this system are fouled easily by suspended solids and oils in liquid radwaste that was flowed in this system. In this study, an experiment to improve quality of water which was flowed in SIES was performed. and design data of Scale-up pretreatment process were secured. Also, each module design for Microfiltration and Nanofiltration unit of the pretreatment process for SIES was performed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.57-57
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• 2018

The electrorefining process is generally composed of two recovery steps in pyroprocessing - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The liquid cathode processing is necessary to separate cadmium from the actinide elements since the actinide deposits are dissolved or precipitated in a liquid cathode. Distillation process was employed for the cathode processing. It is very important to avoid a splattering of cadmium during evaporation due to the high vapor pressure. In this study, a multi-layer porous round cover was proposed and examined to develop a splatter shield for the Cd distillation crucible. Cadmium vapor can be released through the holes of the shield, whereas liquid drops can be collected in the multiple hemisphere. The collected drops flow on the round surface of the cover and flow down into the crucible. The crucible cover was fabricated and tested in the Cd distiller. The cover was made with three stainless steel round plates with a diameter of 33.50 mm. The distance between the hemispheres and the diameter of the holes are 10 and 1 mm, respectively. About 40 grams of Cd and about 4 grams of Bi was distilled at a reduced pressure for two hours at $470^{\circ}C$. After the Cd distillation experiment, cadmium was not detected and more than 90 % of Bi remained in the ICP-OES analysis. Therefore the crucible cover can be a candidate for the splatter shield of the Cd distillation crucible. Further development of the crucible cover is necessary for the decision of the optimum cover geometry and the operating conditions of the Cd distiller.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.101-104
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• 2003

In this study based on the mass transfer theory, experiments for the evaporation rates depending on various conditions were carried out through the operation of the existing Natural Evaporation Facility in KAERI. Evaporation media were made of the cotton and polyester. Air circulation in the facility was forced by exhausting fans. The evaporation rate and the decontamination factor were calculated by the result of experiment. The evaporation rate increased as the flow rate of air supply, the feed rate of liquid waste, and the temperature of supplied air increased. As for the humidity of supplied air, the evaporation rate was getting higher as the humidity was getting lower. As the result of this study, operation conditions of the Natural Evaporation Facility are optimized as follows : The air temperature above $8^{\circ}C$, the air humidity below 70%, the air flow rate 1.14-1.47 m/sec, and the liquid waste feed rate $4.6{\ell}/hr\cdotm^2$. The decontamination factor and the radioactivity are $5.1{\times}10^3$and $4.7{\times}10^{-13}{\mu}Ci/\textrm{m}{\ell}$ respectively, at the above mentioned optimum operation conditions. The air factor in the Dalton's equation for evaporation was determined from results of experiment on the temperature, the humidity, and the flow rate of supplied air as following : $[\textit{Eh}=(0.018 + 0.0141\textitv) {\delta}textitH]$

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.463-469
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• 2003

PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents from PWRs. In PWR-GALE Code, Auxiliary building iodine removal efficiency, one of the code input data, did not reflect adequately the new design of KSNP which has two auxiliary buildings(PAB and SAB). In this study, we developed a revised method how to correct iodine removal efficiency in KSNP. And newly proposed methodology through case study using Ul-Jin 5, 6 design data was verified.