• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.49-56
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• 2010

In the pyroprocessing of spent nuclear fuels, LiCl-KCl waste salt containing radioactive rare earth chlorides are generated. The radioactive rare earth oxides are recovered by co-oxidative precipitation of rare earth elements. The powder phase of rare eath oxide waste must be immobilized to produce a monolithic wasteform suitable for storage and ultimate disposal. The immobilization of these waste developed in this study involves a solid state sintering of the waste with host borosilicate glass and zinc titanate based ceramic matrix(ZIT). And the rare-earth monazite which synthesised by reaction of ammonium di-hydrogen phosphate with the rare earth oxides waste, were immobilzed with the borosilicate glass. It is shown that the developed ZIT ceramic wasteform is highly resistant the leaching process, high density and thermal conductivity.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2889-2896
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• 2023

The waste acceptance criteria for heavy metals in mixed waste should be developed by reflecting the leaching behaviors that could highly depend on the repository design and environment surrounding the waste. The current standards widely used to evaluate the leaching characteristics of heavy metals would not be appropriate for the silo-type repository since they are developed for landfills, which are more common than a silo-type repository. This research aimed to explore the leaching behaviors of cementitious waste with Pb, Cd, and Sb metallic and oxide powders in an environment simulating a silo-type radioactive waste repository. The Toxicity Characteristic Leaching Procedure (TCLP) and the ANS 16.1 standard were employed with standard and two modified solutions: concrete-saturated deionized and underground water. The compositions and elemental distribution of leachates and specimens were analyzed using an inductively coupled plasma optical emission spectrometer (ICP-OES) and energy-dispersive X-ray spectroscopy combined with scanning electron microscopy (SEM-EDS). Lead and antimony demonstrated high leaching levels in the modified leaching solutions, while cadmium exhibited minimal leaching behavior and remained mainly within the cement matrix. The results emphasize the significance of understanding heavy metals' leaching behavior in the repository's geochemical environment, which could accelerate or mitigate the reaction.

• Journal of the Korean Ceramic Society
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• v.52 no.2
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• pp.92-102
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• 2015

Current plans for legacy nuclear wastes stored in underground tanks at the U.S. Department of Energy's Hanford Site in Washington are that they will be separated into high-level waste and low-activity waste fractions that will be vitrified separately. Formulating optimized glass compositions that maximize the waste loading in glass is critical for successful and economical treatment and immobilization of these nuclear wastes. Glass property-composition models have been developed and applied to formulate glass compositions for various objectives for the past several decades. Property models with associated uncertainties combined with composition and property constraints have been used to develop preliminary glass formulation algorithms designed for vitrification process control and waste-form qualification at the planned waste vitrification plant. This paper provides an overview of the current status of glass property-composition models, constraints applicable to Hanford waste vitrification, and glass formulation approaches that have been developed for vitrification of hazardous and highly radioactive wastes stored at the Hanford Site.

• Membrane and Water Treatment
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• v.9 no.5
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• pp.327-334
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• 2018

The nano/micro composites with highly porous surface area have attracted of great interest, particularly the synthesis of porous and thin film sheets of high performance. In this paper, an easy method of cost-effective synthesis of thin film ceramic fiber membranes based on Hydroxyapatite, and activated carbon by turned into studied to be applied within the service-facilitated the transport of radioactive waste such as $^{90}Sr$, $^{137}Cs$ and $^{60}Co$) as activated product of radioisotopes from ETRR-2 research reactor and dissolved in 3M $HNO_3$, across a thin flat-sheet supported liquid membrane (TFSSLM). Radionuclides are transported from alkaline pH values. The presence of sodium salts in the aqueous media improves in $HNO_3$, the lowering of permeability because the initial $HNO_3$ concentration is improved. The study some parameters on the thin sheet ceramic supported liquid membrane. EDTA as stripping phase concentration, time of extraction and temperature were studied. The study of maximum permeability of radioisotopes for all parameters. The pertraction of a radioactive waste solution from nitrate medium were examined at the optimized conditions. Under the optimum experimental 98.6-99.9% of $^{90}Sr$, 79.65-80.3% of $^{137}Cs$ and $^{60}Co$ 45.5-55.5% in 90-110 min with were extracted in 10-30 min, respectively. The process of diffusion in liquid membranes is governed by the chemical diffusion process.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.621-626
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• 2003

A DUPIC nuclear fuel is a newly developed fuel for CANDU reactors based on the concept of refabrication of spent PWR fuel by a dry process. Because a spent PWR fuel, a highly radioactive material, is used as a starting material, the experimental verification of DUPIC nuclear fuel fabrication requires an appropriate facility which should satisfy engineering requirements and guarantees safe operation. DUPIC nuclear fuel development team modified M6 hot-cell in IMEF to construct the dedicated facility(DFDF) for tile experiment. The experiment with spent PWR fuel have been conducted since January of 2000. Environmental effects of DFDF normal operation have been investigated when DUPIC nuclear fuel is fabricated with the maximum capacity of 50kg U/yr. The analysis results of the radiological safety of DFDF facility have shown that both national regulation limit and IMEF design criteria are satisfied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.79-88
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• 2014

We have developed a smart decontamination device for small-size radioactive scrap metal (SSRSM) necessarily generated from nuclear facilities. This is a multi-modal device such as rotation of magnetic field focusing on the region containing the abrasion pins placed around target and ultrasonic cleaner. Additionally, in order to increase the decontamination efficiency we have modified some configuration of the device so that it could work on them evenly and totally. With the Optimal operating for operation of the new device, we tried to decontaminate some various metal selected as a sample during 15 minutes sequentially using each method, magnetic and ultrasonic device. As a result, the range of decontamination factor has been highly increased to 18~56. After decontamination, all samples were found its activity less than background level.

• Nuclear Engineering and Technology
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• v.9 no.2
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• pp.75-81
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• 1977

This experiment has been carried out to determine the pH dependent cation exchange capacity concerning the sorption phenomenon of long-lived radionuclides contained in low-level liquid radioactive waste on various clay minerals. The pH dependent cation exchange capacity determined by Sawhney's method are used to the analysis of sorption phenomenon. About 70 percent of the total cation exchange capacity is contributed by the pH dependent CEC due to the negative charge originated naturally in clays in case of clinoptilolite, vermiculite and sodalite. It is sugested in this test that the high neutral salt CEC, that is, highly charged clays would show good fixation yield. The removal of radionuclides at the pH range more than pH 9 is considered the hydroxide precipitation of metal ion rather than the cation exchange. The Na-clay prepared by the method of successive isomorphic substitution with electrolyte showed a considerable improvement in removal efficiency for the decontamination.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.209-218
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• 2012

A pyroprocess is currently being developed by KAERI to cope with a highly accumulated spent nuclear fuel in Korea. The pyroprocess produces a certain amount of high-level radioactive waste (HLW), which is solidified by a ceramic binder. The produced ceramic waste will be confined in a secure disposal canister and then placed in a deep geologic formation so as not to contaminate human environment. In this paper, the development of a disposal canister was overviewed by discussing mainly its design premises, constitution, manufacturing methods, corrosion resistance in a deep geologic environment, radiation shielding, and structural stability. The disposal canister should be safe from thermal, chemical, mechanical, and biological invasions for a very long time so as not to release any kind of radionuclides.

• International Journal of Naval Architecture and Ocean Engineering
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• v.4 no.2
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• pp.141-150
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• 2012

Ship damage caused by maritime casualties leads to marine pollution and loss of life and property. To prevent serious damage from maritime casualties, several types of safety regulations are applied in ship design. Damage stability regulation is one of the most important safety issues. Designs of ships for long international voyages must comply with these regulations. Current regulations, however, do not consider the characteristics of the operating route of each ship and reflect only ship size and type of cargo. In this paper, a damage safety assessment was undertaken for a ship carrying radioactive waste in actual wave conditions. Damage cases for safety assessment were constructed on the basis of safety regulations and related research results. Hull form, internal arrangement, loading condition and damage condition were modeled for damage safety simulation. The safety simulation was performed and analyzed for 10 damage cases with various wave heights, frequency and angle of attack on an operating route. Based on evaluation results, a design alternative was generated, and it was also simulated. These results confirmed that damage safety analysis is highly important in the design stage in consideration of the operating route characteristics by simulation. Thus a ship designer can improve safety from damage in this manner.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.1-14
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• 2017

For the removal of cesium (Cs) from high radioactive/high salt-laden liquid waste, this study synthesized a highly efficient composite adsorbent (potassium cobalt ferrocyanide (PCFC)-loaded chabazite (CHA)) and evaluated its applicability. The composite adsorbent used CHA, which could accommodate Cs as well as other molecules, as a supporting material and was synthesized by immobilizing the PCFC in the pores of CHA through stepwise impregnation/precipitation with $CoCl_2$ and $K_4Fe(CN)_6$ solutions. When CHA, with average particle size of more than $10{\mu}m$, is used in synthesizing the composite adsorbent, the PCFC particles were immobilized in a stable form. Also, the physical stability of the composite adsorbent was improved by optimizing the washing methodology to increase the purity of the composite adsorbent during the synthesis. The composite adsorbent obtained from the optimal synthesis showed a high adsorption rate of Cs in both fresh water (salt-free condition) and seawater (high-salt condition), and had a relatively high value of distribution coefficient (larger than $10^4mL{\cdot}g^{-1}$) regardless of the salt concentration. Therefore, the composite adsorbent synthesized in this study is an optimized material considering both the high selectivity of PCFC on Cs and the physical stability of CHA. It is proved that this composite adsorbent can remove rapidly Cs contained in high radioactive/high salt-laden liquid waste with high efficiency.