• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.1
• /
• pp.91-99
• /
• 2007

It was intended in this study to investigate the effects of various parameters on the chemical reaction or mass transfer yield in a tubular-type nuclear waste treatment equipment. Since such equipments, as a tubular reactor, multistage solvent extractor, and adsorption column, accompany chemical reaction or mass transfer along the fluid-flowing direction, mathematical modeling for each equipment was carried out first. Then their behaviors of the chemical reaction or mass transfer were predicted through computer simulations. The inherent major parameters for each equipment were chosen and their sensitivities. affecting the reaction or mass transfer yield were analyzed. For the tubular reactor, the effects of axial diffusion coefficient and reaction rate constant on the reaction yield were investigated. As for the multistage solvent extractor, the backmixing of continuous phase and the distribution coefficient between fluid and solvent were considered as the major parameters affecting the extraction yield as well as concentration profiles throughout the axial direction of the extractor. For the adsorption column, the equilibrium constant between fluid and adsorbent surface, and the overall mass transfer coefficient between the two phases were taken as the major factors that affect the adsorption rate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.3
• /
• pp.189-197
• /
• 2007

The solubility of U(VI) hydrolysis products was determined by using a laser-induced breakdown detection (LIBD) technique. The experiments were carried out at uranium concentrations in range from $2{\times}10^{-4}\;M\;to\;4{\times}10^{-6}\;M$, pH values between 3.8 and 7.0, the constant ionic strength of 0.1 M $NaClO_4$ and the temperature of $25.0{\pm}0.1^{\circ}C$. The solubility product of U(VI) hydrolysis products was calculated from LIBD results by using the hydrolysis constants selected in NEA-TDB. The solubility product extrapolated to zero ionic strength, ${\log}K^{\circ}_{sp}=-22.85{\pm}0.23$ was calculated by using a specific ion interaction theory (SIT). The spectral features of ionic species in uranium solutions were investigated by using a conventional UV-visible absorption spectrophotometer and a fluorophotometer, respectively, $(UO_2)_2(OH)_2^{2+}\;and\;(UO_2)_3(OH)_5^+$ were dominant species at uranium concentration of $2{\times}10^{-4}\;M$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.2
• /
• pp.91-101
• /
• 2007

In this study, domestic regulatory requirement was investigated for self-disposal of concrete waste from nuclear fuel processing facility. And after self-disposal as landfill or recycling/reuse, the exposure dose was evaluated by RESRAD Ver. 6.3 and RESRAD BUILD Ver.3.3 computing code for radiological assessments of the general public. Derived clearance level by the result of assessments for the exposure dose of the general public is 0.1071Bq/g (3.5% enriched uranium) for landfill and $0.05515Bq/cm^2$ (5% enriched uranium) for recycling/reuse respectively. Also, residual radioactivity of concrete waste after decontamination was investigated in this study. The result of surface activity is $0.01Bq/cm^2\;for\;{\alpha}-emitter$ and the result of radionuclide analysis for taken concrete samples from surface of concrete waste is 0.0297Bq/g for concentration of $^{238}U$, below 2w/o for enrichment of $^{235}U$ and 0.0089Bq/g for artificial contamination of $^{238}U$ respectively. Therefore, radiological hazard of concrete waste by self-disposal as landfill and recycling/reuse is below clearance level to comply with clearance criterion provided for Notice No.2001-30 of the MOST and Korea Atomic Energy Act.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.2
• /
• pp.123-132
• /
• 2007

This study was performed to evaluate the co- and mutual separation for Am, Cm and RE elements from the simulated multi-component solution equivalent to real HLW level by a Zr-DEHPA(di-(2-ethylhexyl) phosphoric acid containing Zirconium)/$NDD(n-dodecane)-HNO_3$ extraction system. Zr-DEHPA was self-synthesized and the optimal condition of (15g/L Zr-1M DEHPA)/NDD-1M $HNO_3$ was selected taking into consideration of prevention of the third phase, and effects of concentration of DEHPA, nitric acid and impregnant amount of Zr on the co-extraction of Am, Cm and RE. In that condition, the extraction yields were 81% (Am), 85% (Cm), more than 80% (RE elements), 98% (Mo), 85% (Fe), 98% (U), 73% (Np), and less than 5% (other elements) so that the system developed for the co-extraction of Am-Cm/RE was proved to be available. For that, however, U, Np, Mo and Fe was elucidated to have to be removed in advance, and Zr inducing the third phase formation was found to be practically excluded. The co-extracted Am-Cm/RE were sequentially separated in an order of Am-Cm (stripping agent : 0.05 M DTPA-1M Lactic acid of pH 3.6)${\rightarrow}RE$ (stripping agent : 5M $HNO_3$), and then their separation factors were evaluated. At above conditions, Am of 65.4%, Cm of 63.9%, RE (except for Y) of more than 85% were stripped.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.2
• /
• pp.133-143
• /
• 2007

This study was performed to evaluate the co- and sequential separation of Tc, Np and U from the simulated multi-component HLW solution by a TBP (tributyl phosphate)-TOA (tri- octyl amine)/NDD $(n-dodecane)-HNO_3$ extraction system. An optimal condition of (30% TBP-0.5% TOA)/NDD-1 M $HNO_3$ was selected by taking account of a prevention of the 3rd phase and effects of concentration of TBP, TOA and nitric acid on the co-extraction of Tc, Np and U. In that condition, the extraction yields were 81% (Tc), 85% (Np), less than 9% (Am and RE elements), about 8% (Pd), and less than 5% (other elements) so that the system developed for the co-extraction of Tc, Np and U was proved to be available. For that, however, more than 99% of Zr was found to be pre-removed. The co-extracted Tc, Np and U were sequentially separated in order of Tc(stripping agent : 5 M $HNO_3$)${\rightarrow}Np$ by reductive stripping (reductive-stripping agent : 0.1 M AHA)${\rightarrow}U$ (stripping agent : 0.01 M $HNO_3$), and then their separation factors were evaluated. At these conditions, 95% of Tc, 98% of Np and 99% of U could be recovered in each step.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.3
• /
• pp.347-357
• /
• 2018

The applicability of room temperature ionic liquids (RTILs), 1-alkyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([$C_nmim$] [$Tf_2N$]), was investigated for the extraction of Am(III) and Eu(III) from nitric acid using n-octyl(phenyl)-N,N-diisobutyl carbamoylmethyl phosphine oxide (CMPO) and tri-n-butylphosphate (TBP) as extractants. The distribution ratios of Am(III) and Eu(III) in CMPO-TBP/[$C_nmim$][$Tf_2N$] were measured as a function of various parameters such as the concentrations of nitric acid, CMPO, and TBP. The results were compared with those obtained in CMPO-TBP/n-dodecane (n-DD). With comparable concentrations of the extractants, the distribution ratios obtained with RTILs were much higher than those obtained with n-DD. It was observed that the extraction efficiency was less for Eu(III) than for Am(III). The extraction of Am(III) and Eu(III) decreased with increases in the feed acidity for all three RTILs. The results suggest that the extraction of Am(III) and Eu(III) by CMPO in RTILs from nitric acid proceeds through the cation-exchange mechanism. The distribution ratios of Am(III) and Eu(III) increased with increases in the concentration of CMPO for all three RTILs. A linear regression analysis of the extraction data resulted in a straight line with a slope of about 3, suggesting the involvement of 3 molecules of CMPO during the extraction process.

• Analytical Science and Technology
• /
• v.26 no.1
• /
• pp.86-90
• /
• 2013

$^{222}Rn$ concentrations in the groundwater samples without standard material due to the short half-life (3.82 day) were measured through the establishment of the counting efficiency of LSC (Liquid Scintillation Counter) using a standard source of $^{226}Ra$. This study for Quality Assurance/Quality Control (QA/QC) of $^{222}Rn$ analysis was performed to analyze blank samples, duplicate samples, samples of groundwater sampling before and after. In-situ blank samples collected were in the range of 0.44~6.28 pCi/L and laboratory samples were in the range of 1.66~4.95 pCi/L. Their correlation coefficient was 0.9691 and the source contamination from sampling, migration and keeping of samples were not identified. The correlation coefficient between original and duplicate samples from 65 areas was 0.9987. Because radon is an inert gas, in case of groundwater sampling, it is considered to affect the radon concentration. We analyzed samples separately by groundwater sampling before and after using distilled water, but there is no significant difference for $^{222}Rn$ concentrations in distilled waters of two types.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.2
• /
• pp.161-170
• /
• 2006

PFC(Perfluorocarbon) decontamination process is one of best methods to remove hot particulate adhered on the inner surface of hot cell and surface of equipment in hot cell. It was necessary to develop a filtration equipment to reuse the PFC waste solution generated on PFC decontamination due to the high cost of PFC solution and for minimization of the volume of second waste solution. The filtration equipment was developed to remove hot particulate in PFC waste solution. It was made suitable size and weight in consideration of hot cell gate and crane. And it has wheels for easy movement. Flux of the filtration equipment decreased with particulate concentration increase. It consists of pre-filter($1.4{\mu}m$) and final-filter($0.2{\mu}m$) for protection of the flux decrease along filtration time. It treatment capacity of waste solution is 0.2 L/min.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.4
• /
• pp.207-212
• /
• 2009

In this study, a thermal behavior of $PrCl_3$ as one of the lanthanide chlorides in LiCl-KCl molten salts was investigated in an oxidizing condition. First, a thermo-gravimetric analysis (TGA) of $PrCl_3$ was carried out by an injection of $O_2$ gas. Based on the results, an oxidation of $PrCl_3$ in the molten salts was performed by sparging $O_2$ gas with changing temperatures. According to the TGA data of $PrCl_3$, a dissociation of $PrCl_3$ occurred rapidly by about $380^{\circ}C$ and a conversion of $PrCl_3$ to $PrCl_3$ was completed at about $600^{\circ}C$. The thermal behavior of $PrCl_3$ in LiCl-KCl molten salts by sparging $O_2$ gas was similar to that of $PrCl_3$ in the TGA test, and PrOCl as a insoluble compound in the molten salts was precipitated into the bottom of the molten salts. A conversion of $PrCl_3$ to PrOCl in the molten salts occurred actively at a higher temperature than $650^{\circ}C$. And it would be possible to estimate a conversion status of $PrCl_3$ to PrOCl by measuring a $Cl_2$ concentration in a flue gas generated from an oxidation test of $PrCl_3$ in LiCl-KCl molten salts.

• Journal of Radiation Protection and Research
• /
• v.32 no.1
• /
• pp.1-8
• /
• 2007

A model for radiological dose assessment in an urban environment, METRO-K has been developed. Characteristics of the model are as follows ; 1) mathematical structures are simple (i.e. simplified input parameters) and easy to understand due to get the results by analytical methods using experimental and empirical data, 2) complex urban environment can easily be made up using only 5 types of basic surfaces, 3) various remediation measures can be applied to different surfaces by evaluating the exposure doses contributing from each contamination surface. Exposure doses contributing from each contamination surface at a particular location of a receptor were evaluated using the data library of kerma values as a function of gamma energy and contamination surface. A kerma data library was prepared fur 7 representative types of Korean urban buildings by extending those data given for 4 representative types of European urban buildings. Initial input data are daily radionuclide concentration in air and precipitation, and fraction of chemical type. Final outputs are absorbed dose rate in air contributing from the basic surfaces as a function of time following a radionuclide deposition, and exposure dose rate contributing from various surfaces constituting the urban environment at a particular location of a receptor. As the result of a contaminative scenario for an apartment built-up area, exposure dose rates show a distinct difference for surrounding environment as well as locations of a receptor.