• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1717-1722
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• 2021

This study is devoted to the ability of carbon fiber material samples impregnated with various amounts of barium iodide and triethylenediamine to remove radioactive methyliodide from air streams. The main sorption characteristics of impregnated UVIS® carbon fiber were determined and the use of this material for purifying of technological gas flows at nuclear power plants was evaluated. The methyliodide trapping efficiency by samples impregnated with barium iodide, TEDA, and their mixture was 83.4 ± 0.8%; 93.1 ± 0.6% and 93.5 ± 0.7% respectively, under the same conditions. The study established a significantly higher capacity (8.3 ± 0.07 mg/cm²) of samples impregnated simultaneously with both chemical compounds toward methyliodide. Under the same test conditions, the values of this parameter for the samples impregnated separately with TEDA and BaI₂ were 2.85 ± 0.05 mg/cm² and 0.86 ± 0.04 mg/cm², respectively.

• Clean Technology
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• v.14 no.2
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• pp.110-116
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• 2008

The radioactive wastes generated from the nuclear industry can be divided into the forms of solid, liquid, or gas. Radioactive methyl iodide, a gaseous radioactive waste, is absorbed by activated carbon with 5 wt% of Trietylenediamine (1,4-diazania-bicycle[2.2.2]octane, TEDA) impregnated on the surface. Methyl Iodide ($CH_3I$) is combined chemically with TEDA (the final product : I-TEDA). To recycle radioactive activated carbon, removal of I-TEDA from activated carbon is needed. A wet method for recycling impregnated active carbon was developed to remove radioactive I-TEDA using an acetonitrile solution, which produces lots of secondary wastes. We suggest the removal of I-TEDA by supercritical carbon dioxide with co-solvents. In this experiment, we used a quartz crystal microbalance (QCM) for measuring the removal rate of the I-TEDA. From the experimental results, methanol was found to be the optimum co-solvent, and the optimum conditions such as temperature, pressure, and co-solvent flow rate were obtained. Possibility of using supercritical fluid in the removal of I-TEDA from radioactive activated carbon was also discussed.

• Proceedings of the IEEK Conference
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• 2001.10a
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• pp.586-591
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• 2001

This research investigated the feasibility of the accelerated carbonation of cement waste forms with carbon dioxide in a supercritical state. Hydraulic cement has been used as a main solidification matrix for the immobilization of radioactive and/or hazardous wastes. As a result of the hydration reaction for major compounds of portland cement, portlandite (Ca(OH)$_2$) is present in the hydrated cement waste form. The chemical durability of a cement form is expected to increase by converting portlandite to the less soluble calcite (CaCO$_3$). For a faster reaction of portlandite with carbon dioxide, SCCD (supercritical carbon dioxide) rather than gaseous $CO_2$, in ambient pressure is used. The cement forms fabricated with an addition of slated lime or Na-bentonite were cured under ambient conditions for 28days and then treated with SCCD in an autoclave maintained at 34$^{\circ}C$ and 80atm. After SCCD treatment, the physicochemical properties of cement matrices were analyzed to evaluate the effectiveness of accelerated carbonation reaction. Conversion of parts of portlandite to calcite by the carbonation reaction with SCCD was verified by XRD (X-ray diffraction) analysis and the composition of portlandite and calcite was estimated using thermogravimetric (TG) data. After SCCD treatment, tile cement density slightly increased by about 1.5% regardless of the SCCD treatment time. The leaching behavior of cement, tested in accordance with an ISO leach test method at 7$0^{\circ}C$ for over 300 days, showed a proportional relationship to the square root of the leaching time, so the major leaching mechanism of cement matrix was diffusion controlled. The cumulative fraction leached (CFL) of calcium decreased by more than 50% after SCCD treatment. It might be concluded that the enhancement of the characteristics of a cement matrix by an accelerated carbonation reaction with SCCD is possible to some extent.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.193-200
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• 2005

[ $^{14}C$ ] and $^{3}H$ in the evaporator bottom (EB) discharged from the Nuclear power plant (NPP) were extracted simultaneously into a gaseous $^{14}CO_{2}$ and liquefied HTO by using the chemical oxidation, which is the method to oxidize samples completely using potassium persulfate and sulfuric acid. The extracted $^{14}C$ and $^{3}H$ were counted by the liquid scintillation counter (LSC) after the quench correction. To examine the recovery of $^{14}C$ using the radioactive standards, $Na_{2}^{14}CO_{3}$, $^{14}C-alcohol$, and $^{14}C-toluene$ as $^{14}C$, and HTO as $^{3}H$ were used. Also, the most suitable method for oxidizing $^{14}C-toluene$, which is difficult to be oxidized, was investigated through FT-IR spectra according to the concentration of sulfuric acid. With the identical method, $^{14}C$ and $^{3}H$ in the EB generated in the NPP were assayed in the range of $8.35{\sim}l.38{\times}10^3$ Bq/g and $2.46{\times}10^2{\sim}1.40{\times}10^4$ Bq/g, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.107-120
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• 2019

The decommissioning of one nuclear power plant in a multi-unit nuclear power plant (multi-unit NPP) site may pose radiation exposure risk to decommissioning workers. Thus, it is essentially required to evaluate the exposure dose of decommissioning workers of operating multi-unit NPPs nearby. The ENDOS program is a dose evaluation code developed by the Korea Atomic Energy Research Institute (KAERI). As two sub-programs of ENDOS, ENDOS-ATM to anticipate atmospheric transport and ENDOS-G to calculate exposure dose by gaseous radioactive effluents are used in this study. As a result, the annual maximum individual dose for decommissioning workers is estimated to be $2.31{\times}10^{-3}mSv{\cdot}y^{-1}$, which is insignificant compared with the effective dose limit of $1mSv{\cdot}y^{-1}$ for the public. Although it is revealed that the exposure dose of operating multi-unit NPPs does not result in a significant impact on decommissioning workers, closer examination of the effect of additional exposure due to actual demolition work is required. The calculation method of this study is expected to be utilized in the future for planned decommissioning projects in Korea. Because domestic NPPs are located in multi-unit sites, similar situations may occur.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1517-1523
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• 2020

The widely used medical isotope technetium-99 m (^99mTc) is a daughter of Molybdenum-99 (⁹⁹Mo), which is mainly produced using dedicated research reactors from the nuclear fission of uranium-235 (²³⁵U). ^99mTc has been used for several decades, which covers about 80% of the all the nuclear diagnostics procedures. Recently, the instability of the supply has become an important topic throughout the international radioisotope communities. The aging of major ⁹⁹Mo production reactors has also caused frequent shutdowns. It has triggered movements to establish new research reactors for ⁹⁹Mo production, as well as the development of various ⁹⁹Mo production technologies. In this context, a new research reactor project was launched in 2012 in Korea. At the same time, the development of fission-based ⁹⁹Mo production process was initiated by Korea Atomic Energy Research Institute (KAERI) in 2012 in order to be implemented by the new research reactor. The KAERI process is based on the caustic dissolution of plate-type LEU (low enriched uranium) dispersion targets, followed by the separation and purification using a series of columns. The development of proper waste treatment technologies for the gaseous, liquid, and solid radioactive wastes also took place. The first stage of this process development was completed in 2018. In this paper, the results of the hot test production of fission ⁹⁹Mo using HANARO, KAERI's 30 MW research reactor, was described.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.3
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• pp.235-242
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• 2015

A commercially available copper mesh was investigated as an iodine off-gas capturing medium for pyroprocessing, with an aim to replace costly silver based adsorbents. Theoretical calculation results suggested that the reaction between metallic copper and gaseous iodine will occur spontaneously to produce copper iodide in the temperature range of 100 ~ 500℃. The effect of the reaction temperature on iodine capturing efficiency was investigated by experimentation, and it was found that 5 and 6 wt% of iodine (initial mass 2.0 g) was captured by a single copper mesh (0.26 g) at 300 and 400℃, respectively. The repeated experimental results also suggested that copper utilization can be increased with the help of the spontaneous detachment of the reaction product (CuI) from a copper mesh. The formation of the CuI phase was confirmed using the X-ray diffraction technique, and the surface morphology of the reaction product was observed using scanning electron microscopy.