• Proceedings of the Membrane Society of Korea Conference
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• 1997.06a
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• pp.143-158
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• 1997

Electrochemical deionization (EDI) offers continuous demineralization at higher water recovery rates (>90%), compared with mixed bed ion exchange, and without the use of chemical regenerants and the associated production of saline waste water. Although EDI technology has been used in some power generation applications, its wider application requires the satisfactory resolution of outstanding capital cost and performance issues. This paper reports on the field evaluation of a new cost-effective EDI technology in a power generation application. The E-Cell System$^{TM}$, which became commercially available in the fourth quarter of 1996, consists of a rugged, modular system, based on a new high-performance EDI stack. Starting in May 1996, a 100 gpm modular EDI pilot system, rated for operation at 100 psi, was evaluated at the TVA Brown's Ferry Nuclear Plant. The feed consisted of Reverse Osmosis (RO) permeate with a conductivity of 4-7 $\mu$S/cm. The pilot system reliably produced 17.8-18.0 M$\Omega$.cm water under design operating conditions, independent. Silica levels were reduced from ca. 50 ppb to 4 ppb, while TOC levels were reduced from ca. 120 ppb to 30 ppb.

• Applied Chemistry for Engineering
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• v.7 no.1
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• pp.153-161
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• 1996

Several main nuclides($^{241}Am$, $^{152}Eu$ and $^{237}Np$) in radioactive waste solution were selected and examined to mutual separation with di-(2-ethylhexyl) phosphoric acid by solvent extraction technique. $^{237}Np$ was extracted more than 99.9% adding the $H_2O_2$ that was a good reductant for the oxidation state control of $^{237}Np$. $^{241}Am$, $^{152}Eu$ and $^{237}Np$ could be fairly well separated one another during the different sequence stripping stages, but about 7~9.6% of the other nuclides were still remained for the $^{241}Am$ stripping solution. This result shows that the product of $^{152}Eu$ and $^{237}Np$ was good, but $^{241}Am$ may be needed to further purification process. It was also discussed on the cause of the third phase formation phenomenon that was found in the solvent regeneration.

• Journal of Energy Engineering
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• v.21 no.2
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• pp.133-137
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• 2012

Among various clean energy technologies, the solar energy technology has been widely used in various fields such as photovoltaic power generation and solar water/space heating. These days, special attention is drawn on its conversion into acoustic energy along with waste heat as a means to promote clean energy utilization. This work was carried out to investigate the possibility of converting solar energy into acoustic waves, especially, its performance characteristics for a single resonance tube (20.2 mm in ID). Variations are made for the stack length and its position as well as power supply. For a resonance tube of 200mm, an average sound pressure of 114.5 dB was measured with a stack length of 25.6mm at 5cm from the closed end. When the power supply was increased to 35W, an average sound pressure of 117.29 dB was detected with a frequency of 500Hz. There was an increase in frequency, 630 Hz (115.7dB), with a shorter resonance tube of 150mm.

• Analytical Science and Technology
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• v.13 no.5
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• pp.601-607
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• 2000

Lanthanum has been used as one of the burnup monitor in spent nuclear fuel. $U_3Si/Al$ spent nuclear fuel contains small amount of La in high concentration of U and Al. Therefore, chemical separation of La is required to remove matrix elements. At first, ion chromatography (IC) and inductively coupled plasma systems were installed in radiation shielded glove box to handle the radioactive samples. Retention behavior of uranium, aluminum, lanthanum and some interesting fission products (Sr, Zr, Y, Mo, Ru, Pd, Rh, Cs, Ba, Ce, Pr, Nd, Sm, Eu and Cd) was investigated using the CG10 column and ${\alpha}$-HiBA eluent. As all elements were eluted earlier than lanthanum in 0.2 M ${\alpha}$-HiBA eluent, a portion of U and Al was directly passed to waste using a three way valve between the column and the nebulizer. Thus it was possible to determine the lanthanum in a high concentration of U and Al matrix. Retention time of La was about 12 minutes in this separation condition. Optimum range for the determination of La in $U_3Si/Al$ spent nuclear fuel was $1-10{\mu}g/L$ (ppb) with this system and detection limit was $0.25{\mu}g/L$ in case of $200{\mu}L$ of sample volume.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.319-327
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• 2010

A nuclear plant ESF ACS simulator was designed, built, and verified to perform experiment related to ESF ACS of nuclear power plants. The dimension of 3D CAD model was based on drawings of the main control room(MCR) of Yonggwang units 5 and 6. The CFD analysis was performed based on the measurement of the actual flow rate of ESF ACS. The air flowing in ACS was assumed to have $30^{\circ}C$ and uniform flow. The flow rate across the HEPA filter was estimated to be 1.83 m/s based on the MCR ACS flow rate of 12,986 CFM and HEPA filter area of 9 filters having effective area of $610{\times}610mm^2$ each. When MCR ACS was modeled, air flow blocking filter frames were considered for better simulation of the real ACS. In CFD analysis, the air flow rate in the lower part of the active carbon adsorber was simulated separately at higher than 7 m/s to reflect the measured value of 8 m/s. Through the CFD analyses of the ACSes of fuel building emergency ventilation system, emergency core cooling system equipment room ventilation cleanup system, it was confirmed that all three EFS ACSes can be simulated by controlling the flow rate of the simulator. After the CFD analysis, the simulator was built in nuclear grade and its reliability was verified through air flow distribution tests before it was used in main tests. The verification result showed that distribution of the internal flow was uniform except near the filter frames when medium filter was installed. The simulator was used in the tests to confirm the revised contents in Reg. Guide 1.52 (Rev. 3).

• Nuclear Medicine and Molecular Imaging
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• v.41 no.4
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• pp.309-316
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• 2007

Purpose: Sentinel lymph node biopsy became the standard procedure in early breast cancer surgery. Faculty members might be exposed to a trace amount of radiation. The aim of this study is to quantify the radiation exposure and verify the safety of the procedure and the facilities, especially during pathologic process. Materials and Methods: Sentinel lymph node biopsies with Tc-99m human serum albumin were performed as routine clinical work. Exposed radiation doses were measured in pathologic technologist, nuclear medicine technologist, and nuclear medicine physician using a thermoluminescence dosimeter (TLD) during one month. We also measured the residual radioactivities or absorbed dose rates, the exposure distance and time during procedure, the radiation dose of the waste and the ambient equivalent dose of the pathology laboratory. Results: Actual exposed doses were 0.21 and 0.85 (uSv/study) for the whole body and hand of pathology technologist after 47 sentinel node pathologic preparations were performed. Whole body exposed doses of nuclear medicine physician and technologist were 0.2 and 2.3 (uSv/study). According to this data and the exposure threshold of the general population (1 mSv), at least 1100 studies were allowed in pathology technologist. The calculated exposed dose rates (${\mu}$ Sv/study) from residual radioactivities data were 2.47/ 22.4 ${\mu}$ Sv (whole body/hand) for the surgeon; 0.22/ 0 ${\mu}$ Sv for operation nurse. The ambient equivalent dose of the pathology laboratory was 0.02-0.03 mR/hr. The radiation dose of the waste was less than 100 Bq/g and nearly was not detected. Conclusion: Pathologic procedure relating sentinel lymph node biopsy using radioactive colloid is safe in terms of the radiation safety.(Nucl Med Mol Imaging 2007;41(4);309-316)

• The Korean Journal of Nuclear Medicine Technology
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• v.25 no.2
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• pp.41-47
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• 2021

Purpose There are many cyclotrons compared to the land area of the Republic of Korea. Because GMP certification is required and the nuclear medicine test does not apply for insurance, the number of examinations for nuclear medicine is decreasing. Therefore, there is a high probability of early decommissioning of the cyclotron. However, we do not unusually perform the radioactivation evaluation on concrete that can be classified as radioactive waste during the decommissioning of the cyclotron. In this study, we aim to confirm the radioactivation in the concrete surface using Handheld Radionuclide Identification Devices (RIDs). Materials and Methods Because there is no cyclotron being decommissioning in the Republic of Korea, it was impossible to perform the coring of concrete for radioactivation analysis. In this study, we used the KIRAMS-13 and analyzed the concrete surface in the target direction in the cyclotron room. After setting the target direction as the center, radionuclides were measured for about five months at thirty points with vertical and horizontal intervals of 30 cm. We used the RIIDEye(Detector: NaI(Tl) detector, manufacturer: Thermo) in this study and set the measurement time per point to one day (24 hours). Results Co-60 and Cs-137 were detected in some measurement points, and we confirmed the radioactivity of Co-60 detected at the most points. As a result, we found that the radioactivity of Co-60 was high in the diagonal direction (from the lower-left direction to the upper right direction) based on the center of the target. However, we think it is impossible to apply the corresponding results to all cyclotrons because we performed the study using only one cyclotron. Conclusion In thirty measurement points, we could confirm the radioactive nuclides and the relative radioactivity using the results of portable nuclides analyzer. Therefore, we expect that we can use the portable nuclides analyzer to select the coring position of concrete during the decommissioning of the cyclotron. Also, if we secure the radioactivation data for several years, we expect to make a more accurate estimate of radioactive waste during the preparation period of decommissioning of the cyclotron.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.45-53
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• 2012

The metal chloride wastes from a pyrochemical process to recover uranium and transuranic elements has been considered as a problematic waste difficult to apply to a conventional solidification method due to the high volatility and low compatibility with silicate glass. In this study, a dechlorination approach to treat LiCl-KCl waste for final disposal was adapted. In this study, a $SiO_2-Al_2O_3-P_2O_5$ (SAP) inorganic composite as a dechlorination agent was prepared by a conventional sol-gel process. By using a series of SAPs, the dechlorination behavior and consolidation of reaction products were investigated. Different from LiCl waste, the dechlorination reaction occurred mainly at two temperature ranges. The thermogravimetric test indicated that the first reaction range was about $400^{\circ}C$ for LiCl and the second was about $700^{\circ}C$ for KCl. The SAP 1071 (Si/Al/P=1/0.75/1 in molar) was found to be the most favorable SAP as a dechlorination agent under given conditions. The consolidation test revealed that the bulk shape and the densification of consolidated forms depended on the SAP/Salt ratios. The leaching test by PCT-A method was performed to evaluate the durability of consolidated forms. This study provided the basic information on the dechlorination approach. Based on the experimental results, the dechlorination method using a $SiO_2-Al_2O_3-P_2O_5$ (SAP) could be considered as one of alternatives for the immobilization of waste salt.

• Journal of the Korea Concrete Institute
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• v.25 no.3
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• pp.321-329
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• 2013

Recently, the medium-low level radioactive waste from nuclear power plant must be transported from temporary storage to the final repository. Medium-low level radioactive waste, which is composed mainly of the liquid ion exchange resin, has been consolidated with cementitious material in the plastic or iron container. Since cementitious material is brittle, it would generate cracks by impact load during transportation, signifying leakage of radioactive ray. In order to design the safety transporting equipment, there is a need to check the compressive strength of the current waste. However, because it is impossible to measure strength by direct method due to leakage of radioactive ray, we will estimate the strength indirectly by the dynamic modulus of elasticity. Therefore, it must be identified the relationship between of strength and dynamic modulus of elasticity. According to the waste acceptance criteria, the compressive strength of cement based solid is defined as more than 3.44 MPa (500 psi). Compressive strength of the present solid is likely to be significantly higher than this baseline because of continuous hydration of cement during long period. On this background, we have tried to produce the specimens of the 28 day's compressive strength of 3 to 30 MPa having the same material composition as the solid product for the medium-low level radioactive waste, and analyze the relationship between the strength and the dynamic modulus of elasticity. By controling the addition rates of AE agent, we made the mixture containing the ion exchange resin and showing the target compressive strength (3~30 MPa). The dynamic modulus of elasticity of this mixtures is 4.1~10.2 GPa, about 20 GPa lower in the equivalent compressive strength level than that of ordinary concrete, and increasing the discrepancy according to increase strength. The compressive strength and the dynamic modulus of elasticity show the liner relationship.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.101-112
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• 2016

This study investigated the removal of Sr, which was one of the high radioactive nuclides, by adsorption with Barium (Ba) impregnated 4A zeolite (BaA) from high-radioactive seawater waste (HSW). Adsorption of Sr by BaA (BaA-Sr), in the impregnated Ba concentration of above 20.2wt%, was decreased by increasing the impregnated Ba concentration, and the impregnated Ba concentration was suitable at 20.2wt%. The BaA-Sr adsorption was added to the co-precipitation of Sr with $BaSO_4$ precipitation in the adsorption of Sr by 4A (4A-Sr) within BaA. Thus, it was possible to remove Sr more than 99% at m/V (adsorbent weight/solution volume)=5 g/L for BaA and m/V >20 g/L for 4A, respectively, in the Sr concentration of less than 0.2 mg/L (actual concentration level of Sr in HSW). It shows that BaA-Sr adsorption is better than 4A-Sr adsorption in for the removal capacity of Sr per unit gram of adsorbent, and the reduction of the secondary solid waste generation (spent adsorbent etc.). Also, BaA-Sr adsorption was more excellent removal capacity of Sr in the seawater waste than distilled water. Therefore, it seems to be effective for the direct removal of Sr from HSW. On the other hand, the adsorption of Cs by BaA (BaA-Cs) was mainly performed by 4A within BaA. Accordingly, it seems to be little effect of impregnated Ba into BaA. Meanwhile, BaA-Sr adsorption kinetics could be expressed the pseudo-second order rate equation. By increasing the initial Sr concentrations and the ratios of V/m, the adsorption rate constants ($k_2$) were decreased, but the equilibrium adsorption capacities ($q_e$) were increasing. However, with increasing the temperature of solution, $k_2$ was conversely increased, and $q_e$ was decreased. The activation energy of BaA-Sr adsorption was 38 kJ/mol. Thus, the chemical adsorption seems to be dominant rather than physical adsorption, although it is not a chemisorption with strong bonding form.