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  • Author or Editor: J. S. Park x
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Abstract  

The formation property of Mo precipitate was investigated and improved the existing process was using H2O2 that acts as an interfering compound in a subsequent alumina adsorption process. The property of the Mo precipitate was investigated by using SEM, FTIR, TG-DTA, and XRD. The simulated solution consisted of 1M nitric acid containing seven elements (Mo, I, Ru, Zr, Ce, Nd, Sr) and their radioactive tracers. As a result, the precipitate was composed of the Mo precipitate and re-precipitated a-benzoinoxime which was added excessively for increasing the precipitation efficiency. It was confirmed that the Mo precipitate was formed by the reaction of two a-benzoinoxime molecules and one MoO2 2+. Molybdenum precipitate was dissolved in 0.4M NaOH solution within 5 minutes without H2O2. Hydrogen peroxide induced only the rapid dissolution of the a-benzoinoxime re-precipitate. Also, the dissolution method without H2O2 was favorable in the purification aspect because Zr and Ru were contained as a small fraction of 1.3% and 7.7%, respectively, in the dissolving solution.

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The electrochemical reduction of uranium oxide in the treatment of spent nuclear fuel requires a characterization of the LiCl-Li2O salt used as a reaction medium. Physical properties, melting and vaporization are important for the application of the salt and thus they have been investigated by differential scanning calorimetry (DSC) and thermogravimetry (TG), respectively. Experimental data suggest LiCl and Li2O compound formations, leading to a melting point depression of the LiCl and a co-vaporization of the LiCl-Li2O salt.

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Abstract  

It is regarded that the spent resins from the water purification systems of moderator (MOD) and the primary coolant of the Canada deuterium uranium-pressurized heavy water reactor (CANDU-PHWR) are a unique waste, owing to their high 14C and gamma-emitting nuclides. In this work, 14C and 3H contents, anion and cation fractions and the predominant gamma-emitting nuclides of the spent resins from 4 units of CANDUPHWRs, were investigated. Also the chemical species of 14C of the spent resins were determined. For a simultaneous separation of 14C and 3H from the spent resins, the wet oxidation-16 wt% H2SO4 stripping process was utilized. The 14C and 3H activity concentration range of the spent resins of the nuclear power plant (NPP), 4 units of all CANDU-PHWR types, was 2.48E5 Bq/g ∼5.33E6 Bq/g, 1.29E5 Bq/g and ∼2.33E5 Bq/g, respectively. Among the analyzed spent resins, the highest 14C and 3H activity concentration was detected in units 4 and 3, respectively. It was found that more than 92% of the 14C activity concentration was retained on the anion resin and the predominant chemical species was inorganic 14C. It was revealed that the anion resin fraction of the spent resins from unit 1 and unit 2, was about 40% and that of unit 3 and unit 4 was around 60%. More than 80% of the total gamma-radioactivity concentration was associated with the cation fraction of the spent resin. The predominant gamma-emitting nuclide of the spent resin for unit 2 was 137Cs, a fission product, and that for unit 4 was 60Co, a corrosion product.

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This study aimed to develop a chromatographic method to quantitatively determine phenol in fish tissues. This method involves solvent extraction of acidified samples, followed by derivatization to phenyl acetate and analysis with gas chromatography coupled with mass spectrometry (GC–MS). Phenol in a representative tissue sample (belly, gill, or renal tubules), which was homogenized with 2 N sulfuric acid, was extracted with ethyl acetate and derivatized to phenyl acetate using acetic anhydride and K2CO3 in water. An n-butyl acetate extract was injected into the GC–MS. The linearity (r 2) of the calibration curve was greater than 0.996. The analytical repeatability, which is expressed as the relative standard deviation, was less than 6.14%, and the recovery was greater than 96.3%. The method detection limit and the limit of quantitation were 8.0 μg/kg and 26 μg/kg, respectively. The proposed method is also applicable to the analysis of other biological tissues for phenol and its analogs, such as pentachlorophenol.

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Abstract  

It is important to increase a throughput of the salt removal process from uranium deposits which is generated on the solid cathode of electro-refiner in pyroprocess. In this study, it was proposed to increase the throughput of the salt removal process by the separation of the liquid salt prior to the distillation of the LiCl–KCl eutectic salt from the uranium deposits. The feasibility of liquid salt separation was examined by salt separation experiments on a stainless steel sieve. It was found that the amount of salt to be distilled could be reduced by the liquid salt separation prior to the salt distillation. The residual salt remained in the deposits after the liquid salt separation was successfully removed further by the vacuum distillation. It was concluded that the combination of a liquid salt separation and a vacuum distillation is an effective route for the achievement of a high throughput performance in the salt separation process.

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Abstract  

In order to develop an 125I seed for brachytherapy of prostate cancer, a carrier body consisting of Al2O3 and silver powder was developed. To optimize the adsorption conditions of 125I on the rods, various experiments were performed. The adsorption capacity was more than 95% after 4 hours at a volume of 50 μl containing about 5 mCi of 125I. Dosimetric properties were measured for the radial and longitudinal directions. Variations were below 11% in the longitudinal distribution and 5% in the radial distribution. This method is effective for the preparation of 125I seeds to be used in brachytherapy treatment.

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Abstract  

The Korea Atomic Energy Research Institute (KAERI) has launched a decommissioning program of the uranium conversion plant. The sludge waste, which was generated during the operation of the plant and stored in the lagoon, was characterized for the development of the treatment process. The physical properties were measured and chemical compositions and radiological properties analyzed. The main compounds of the sludge were ammonium nitrate, sodium nitrate, calcium nitrate, and calcium carbonate. All heavy radioactive elements such as uranium, thorium and 226Ra were precipitated and deposited at the bottom, and were not dissolved in the concentrated nitrate solution. A possible flow-scheme for processing is presented.

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Abstract  

The thermal degradation and thermal stability of rice husk flour (RHF) filled polypropylene (PP) and high-density polyethylene (HDPE) composites in a nitrogen atmosphere were studied using thermogravimetric analysis. The thermal stability of pure PP and HDPE was found to be higher than that of wood flour (WF) and RHF. As the content of RHF increased, the thermal stability of the composites decreased and the ash content increased. The activation energy of the RHF filled PP composites increased slowly in the initial stage until α=0.3 (30% of thermal degradation region) and thereafter remained almost constant, whereas that of the RHF filled HDPE composites decreased at between 30 and 40 mass% of RHF content. The activation energy of the composites was found to depend on the dispersion and interfacial adhesion of RHF in the PP and HDPE matrix polymers.

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Summary  

In order to evaluate the use of gamma-ray treatment as a pretreatment to conventional biological methods, the effects of gamma-irradiation on biodegradability (BOD5/COD) of textile and pulp wastewaters were investigated. For all wastewaters studied in this work, the efficiency of treatment based on TOC removal was insignificant even at an absorbed dose of 20 kGy. However, the change of biodegradability was noticeable and largely dependent on the chemical property of wastewaters and the absorbed dose of gamma-rays. For textile wastewaters, gamma-ray treatment increased the biodegradability of desizing effluent due to degradation of polymeric sizing agents such as polyvinyl alcohol. Interestingly, the weight-loss showed the highest value of 0.97 at a relatively low dose of 1 kGy. This may be caused by the degradation of less biodegradable ethylene glycol prior to terephthalic acid decomposition. For pulp wastewater, the gamma-ray treatment did not improve the biodegradability of cooking and bleaching of C/D effluents. However, the biodegradability of bleaching E1 and final effluents was abruptly increased up to 5 kGy then slowly decreased as the absorbed dose was increased. The initial increase of biodegradability may be induced by the decomposition of refractory organic compounds such as chlorophenols, which are known to be the main components of bleaching C/D and final effluents.

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Summary  

It is impossible to detect 14C and 3H by direct methods such as γ-spectroscopy because they are pure b-emitters and thus they are classified as hard to measure nuclides (HTM). In this paper the analysis results of 14C and 3H in the low level radioactive wastes (LLWs), including spent ion exchange resin, evaporated bottom and sludge are presented. The LLWs were generated by three nuclear power plants (NPPs), in Korea all with pressurized water type reactors (PWRs). A simultaneous separation procedure for 14C and 3H in LLWs was established by wet oxidation-acid stripping. A liquid scintillation analyzer was used for the measurement of 14C and 3H. It was found that the recovery of 14C and 3H was 82-99 and 78-103%, respectively, by wet oxidation-acid stripping with diluted standard solutions. At the lowest injection of 14C and 3H, i.e., at 1.44 Bq for 14C and 1.22 Bq for 3H, the minimum detectable activity (MDA) of 14C and 3H was calculated as 0.88 and 0.78 Bq/g, respectively, for the minimum allowable sample weight, using wet oxidation and 16 wt% H2SO4 acid. By the wet oxidation-16 wt% H2SO4 stripping method no interfering nuclides were detected in the trapping solution of 14CO2 and the distillate of 3H. The activity concentration range of 14C in the analyzed samples, i.e., spent ion exchange resin, evaporated bottom and sludge, was 0.17-110,000, 8.4-1380 and 0.1-10,006 Bq/g, respectively, and that of 3H in the same was from no detectable to 769, 134-14,383 and 0.7-4820 Bq/g, respectively.

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