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We herein present a suquential procedure for separation of nuclides such as 55Fe, 94Nb, 59,63Ni, 90Sr, and 99Tc from activated carbon and stainless steel samples orignating from low and intermediate level of radioactive waste. Initially, iron was removed through the precipitation of Fe(OH)3 upon the addition of NH4OH. The recoveries of Fe and Nb in the precipitate were 98.4 ± 2.9 and 98.5 ± 0.52%, respectively, while those of Ni, Re, and Sr in the supernatant were 74.0 ± 4.0, 96.7 ± 1.2, and 79.6 ± 1.2%, respectively. However, due to the especially low recovery of Re from the supernatant following anion exchange chromatography, the Fe precipitate was removed after the separation of Re by anion exchange chromatography. Our results indicate that this procedure is suitable for providing pure nuclides from waste samples to allow the measurement of their radioactive activities.

A radiological safety assessment study was performed for the transportation of low level radioactive wastes which are temporarily stored in Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea. We considered two kinds of wastes: (1) operation wastes generated from the routine operation of facilities; and (2) decommissioning wastes generated from the decommissioning of a research reactor in KAERI. The important part of the radiological safety assessment is related to the exposure dose assessment for the incident-free (normal) transportation of wastes, i.e., the radiation exposure of transport personnel, radiation workers for loading and unloading of radioactive waste drums, and the general public. The effective doses were estimated based on the detailed information on the transportation plan and on the radiological characteristics of waste packages. We also estimated radiological risks and the effective doses for the general public resulting from accidents such as an impact and a fire caused by the impact during the transportation. According to the results, the effective doses for transport personnel, radiation workers, and the general public are far below the regulatory limits. Therefore, we can secure safety from the viewpoint of radiological safety for all situations during the transportation of radioactive wastes which have been stored temporarily in KAERI.

One of the promising future nuclear energy systems, the Sodium-cooled Fast Reactor (SFR) has been actively developed internationally. Recently, to improve safety and economics of a SFR further, coupling supercritical CO2 power cycle was suggested. However, there can be a chemical reaction between sodium and CO2 at high temperature (more than 400 °C) when the pressure boundary fails in a sodium–CO2 heat exchanger. To ensure the performance of such a system, it is important to employ a cleaning agent to recover the system back to normal condition after the reaction. When sodium and CO2 react, solid and gaseous reaction products such as sodium carbonate (Na2CO3) and carbon monoxide (CO) appear. Since most of solid reaction products are hard and can deteriorate system performance, quick removal of solid reaction products becomes very important for economic performance of the system. Thus, the authors propose the conceptual method to remove the byproducts with a chemical reaction at high temperature. The chemical reaction will take place between the reaction byproducts and a cleaning agent while the cleaning agent is inert with sodium. Thus, various sodium-based compounds were first investigated and three candidate substances satisfying several criteria were selected; sodium bromate (NaBrO3), sodium chlorate (NaClO3), and sodium tetrafluoroborate (NaBF4). The selected substances were thermally analyzed with the TG/DTA studies. Unfortunately, it was revealed that all candidate substances did not react with Na2CO3 and decomposed before 600 °C. However, since no study has been performed on the issue of cleaning byproducts of Na–CO2 reaction so far, this study provides the basic guideline for the future study and suggests the future research direction and preferred characteristics of cleaning agent of sodium CO2 reaction byproducts.

A procedure for the rapid separation of 99Tc, 90Sr, 55Fe, 94Nb, and 59,63Ni in radioactive waste samples was developed for measuring their activity. After separating 99Tc on an anion-exchange resin column, 90Sr was separated as 90Sr(CaC2O4) co-precipitates, and 59,63Ni, 55Fe, and 94Nb were sequentially separated on an anion-exchange resin column in the presence of oxalate. The separated 90Sr and 59,63Ni were further purified on extraction chromatographic resin columns, respectively. The chemical yields of the separation procedure for Re, Sr, Fe, Nb, and Ni, which were added as chemical carriers, are 99.3 ± 0.3 %, 94.8 ± 0.6 %, 93.3 ± 1.2 %, 95.3 ± 1.9 %, and 92.6 ± 0.6 %, respectively. The separation of these radionuclides from sixteen radioactive waste samples required only seven person-days.

We developed a rapid quantification method of Pu, Am, and Cm in low- and intermediate-level radioactive waste using TRU resin, alpha spectrometry, and thermal ionization mass spectrometry for radioactive waste disposal. In particular, we focused on the optimization of sample preparation methods for alpha spectrometry and mass spectrometry to reduce the analysis time. The optimized method used Am and Pu stripping solutions for micro-coprecipitation and showed good recovery and alpha spectrum resolution for alpha counting. In addition, appropriate selection of Pu elution solution decreased the sample preparation time for thermal ionization mass spectrometry measurements.

Surface modified carbon felts were utilized as an electrode for the removal of inorganic ions from seawater. The surfaces of the carbon felts were chemically modified by alkaline and acidic solutions, respectively. The potassium hydroxide (KOH) modified carbon felt exhibited high Brunauer-Emmett-Teller (BET) surface areas and large pore volume, and oxygencontaining functional groups were increased during KOH chemical modification. However, the BET surface area significantly decreased by nitric acid (HNO3) chemical modification due to severe chemical dissolution of the pore structure. The capability of electrosorption by an electrical double-layer and the efficiency of capacitive deionization (CDI) thus showed the greatest enhancement by chemical KOH modification due to the appropriate increase of carboxyl and hydroxyl functional groups and the enlargement of the specific surface area.

We have measured the electronic absorption spectra of the U(III) ion in LiCl–KCl eutectic melt at 450 °C to understand its chemical behavior in the context of pyrochemical process of spent nuclear fuel. The UV–VIS spectra of the U(III) ion consist of two main peaks in the range of 400–600 nm which are attributable to the 5f 3 –5f 2 6d 1 transitions. With the aid of UV–VIS spectroscopic tool, in-situ measurement of chemical reactions of the U(III) with oxide ion as well as neodymium oxide was successfully achieved. The U(III) ion forms insoluble uranium oxide phases by reacting with oxide ion and lanthanide oxides.

In order to evaluate radionuclide inventories as an essential item for the permanent disposal of spent fuel storage racks, chemical conditions for a sample pretreatment of a spent fuel storage rack were studied. Especially, the surface microstructure and the radionuclide distributions for the spent fuel storage rack were investigated by using a SEM–EDX and γ-spectrometer for minimizing the matrix effect which could affect a chemical separation process of some β-emitting radionuclides. The samples were pretreated with a mixed solution of 5 M HCl and 2 M HNO3 by an ultrasonic surface leaching method. Some radionuclides in the raw racks showed the radioactivity of 102–103 Bq for about 10 g of sample weight. From the sample pretreatment, it was confirmed that almost all radionuclides in the rack were completely extracted from the rack when the dissolved thickness of the rack became a maximum 15 μm by the ultrasonic surface leaching method. The established pretreatment method was applied for all spent fuel storage rack generated from Korean NPPs to determine the scaling factor. The radioactivities of 60Co and 137Cs radionuclides in the pretreated solutions were in the range of 4.9E−1~1.5E+2 and 1.2E−1~9.0E+0 Bq/g, respectively.

A hydrous oxide film for the application as dielectric film is synthesized by immersion of pure aluminum in hot water. From a Rutherford backscattering analysis, the ratio of aluminum to oxygen atoms was found to be 3:2 in the anodized aluminum oxide film, and 2:1 in the hydrous oxide layer. Anodization of the hydrous oxide layer was more effective for the transition of amorphous anodic oxides to the crystalline aluminum oxides.

The electrochemical properties on capacitive deionization (CDI) of carbon cloth in NaCl solution have been investigated. For KOH etched carbon cloth, hydroxyl and carbonyl functional groups have increased significantly, while carboxyl functional group has changed little with etching time. For HNO 3 treated carbon surface, carboxyl, carbonyl, and hydroxyl functional groups were produced considerably, and then slightly formed with etching time. From the evaluations of CDI process the carbon cloth electrodes etched in KOH or HNO 3 for 3 h represent higher kinetic constants than that for the non-treated carbon cloth.

We investigated the effects of additives to the etching solution of 1 M hydrochloric acid on the electrochemical etching behavior for aluminum electrolytic capacitors, using scanning and transmission electron microscopy, and AC impedance spectroscopy. For the addition of 1 M sulfuric acid or 5% ethylene glycol to the hydrochloric acid solution, the distribution of etch tunnels was more uniform with high density of etch pits compared with that without addition. The highest specific surface area was obtained from the electrolyte with 5% ethylene glycol additive. The correlation of internal morphologies of etched foils with impedance parameters was interpreted by impedance techniques.

The characteristics and growth behaviors of alumina dielectric layer formed by anodic oxidation were investigated. The aluminum oxide layer anodized at 400 V was predominantly amorphous alumina, but at the applied potentials more than 500 V, amorphous and crystalline γ-alumina were existed in anodic oxide layer and the ratio of γ-alumina increased with the increasing applied potential. During the heat treatment at 600∘C or higher temperature, amorphous alumina layer was transformed into the crystalline γ-alumina. The phase transition of anodic amorphous alumina into crystalline depends on anodic applied potentials and heat-treatment temperatures.

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.

Activated nanostructured-carbon cloths with a high ratio of surface area to volume are used as electrode for capacitive deionization. The electrochemical properties on capacitive deionization for NaCl solution have been investigated to improve efficiency of capacitive deionization properties from aqueous solution, employing chemical surface-modification by etching in alkaline and acidic solution. The removal efficiency of inorganic salts of activated carbon cloths by chemical modification significantly increased. Specially the carbon cloth surface modified in HNO3 showed an effect of improvement in the CDI efficiency due to not only ion adsorption by an electric double layer, but also electron transfer by Faradaic reaction.

In the original publication, the units of the activities in Tables 2 and 3 are incorrect and they need to be corrected to Bq/kg. In addition, ‘Activity of Am’ in the right column in Table 3 needs to be corrected to ‘Activity’. Therefore, corrected Tables 2 and 3 are given below: Table 2 Activity measurement of Am and Cm isotopes in the radioactive dry waste sample using alpha spectrometry Sample # Activity (Bq/kg) Am recovery (%) Am Cm Cm

The concentrations of several radionuclides in low and intermediate level radioactive waste (LILW) drums have to be determined before shipping to disposal facilities. A notice, by the Ministry of Science and Technology (MOST) of the Korean Government, related to the disposal of LILW drums came into effect at the beginning of 2005, with regards to a radionuclide regulation inside a waste drum. MOST allows for an indirect radionuclide assay using a scaling factor to measure the inventories due to the difficulty of nondestructively measuring the essential {alpha} and {beta}-emitting nuclides inside a drum. That is, a scaling factor calculated through a correlation of the {alpha} or {beta}-emitting nuclide (DTM, Difficult-To-Measure) with a {gamma}-emitting nuclide (ETM, Easy-To-Measure) which has systematically similar properties with DTM nuclides. In this study, radioactive wastes, such as spent resin and dry active waste which were generated at different sites of a PWR and a site of a PHWR type Korean NPP, were partially sampled and analyzed for regulated radionuclides by using radiochemical methods. According to a reactor type and a waste form, the analysis results of each radionuclide were classified. Korean radwaste scaling factor was derived from database of radionuclide concentrations. (authors)

The oxidation studies of a sulfur to a sulfate ion by various oxy-halide oxidants in organic (thiourea, methionine) and inorganic (sulfate, thiophosphate) compounds were carried out in an acidic solution. The optimized result of the oxidation reaction was obtained when a bromate compound (BrO{sub 3}{sup -}) as an oxidant and a 3 M HNO{sub 3} solvent. The chemical yield for the oxidation of the organic and inorganic sulfur compounds to a sulfate ion was monitored as 80% for thiophosphate, 87% for methionine, and 100% for thiourea and sulfate within 5% RSD. The oxidation of thiourea required at least 1.6 equivalents of the bromate in an acidic solution. In the case of the oxidation of methionine and thiophosphate, the oxidation yield was above 80% if the bromate was used at 20 times that of the substrates. The chemical yield in the paper sample (WypAll) exceeded 100% because of its background sulfur contents (910 ppm). The sulfate ion was quantitatively measured by using GPC and/or LSC counting of 3 S followed by precipitates of BaSO{sub 4}. The interfering nuclides ({sup 14}C, {sup 32}P) were removed in an acidic condition. The minimum detectable activity (MDA) of {sup 35}S was found to be 0.1more » Bq/g. (authors)« less

Utilization facilities for radioisotope (RI) are increasing annually in South Korea, and the total number was 2,723, as of December 31, 2005. The inspection of a clearance level is a very important problem in order to ensure a social reliance for releasing radioactive materials to the environment. Korean regulations for such a clearance are described in Notice No. 2001–30 of the Ministry of Science and Technology (MOST) and Notice No. 2002–67 of the Ministry of Commerce, Industry and Energy (MOCIE). Most unsealed sources in RI waste drums at a storage facility are low-level beta-emitters with short half-lives, so it is impossible to measure their inventories by a nondestructive analysis. Furthermore, RI wastes generated from hospital, educational and research institutes and industry have a heterogeneous, multiple, irregular, and a small quantity of a waste stream. This study addresses a representative (master) sampling survey and analysis plan for RI wastes because a complete enumeration of waste drums is impossible and not desirable in terms of a cost and efficiency. The existing approaches to a representative sampling include a judgmental, simple random, stratified random, systematic grid, systematic random, composite, and adaptive sampling. A representative sampling plan may combine two or more of the above sampling approaches depending on the type and distribution of a waste stream. Stratified random sampling (constrained randomization) is proven to be adequate for a sampling design of a RI waste regarding a half-life, surface dose, undertaking time to a storage facility, and type of waste. The developed sampling protocol includes estimating the number of drums within a waste stream, estimating the number of samples, and a confirmation of the required number of samples. The statistical process control for a quality assurance plan includes control charts and an upper control limit (UCL) of 95% to determine whether a clearance level is met or not.

Synthesis of U 3+ with Cd 2+ or Bi 3+ was investigated in a LiCl―KCl eutectic melt at 500°C by using cyclic voltammetry (CV) and ultraviolet-visible (UV-vis) spectroscopy. CV results showed that the equilibrium potentials of the species are larger in sequence of U 0/3+ < Cd 0/2+ < U 3+/4+ < Bi 0+/3+ . UV-vis spectroscopy results exhibited that U 3+ was directly produced in the eutectic melt containing Cd 2+ . In BiCl 3 , U metal was oxidized to U 4+ by Bi 3+ , and then U 4+ was reduced to U 3+ by additional U. From these results, the reaction mechanisms taking place during the synthesis are proposed.