• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.05a
• /
• pp.35-36
• /
• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.10a
• /
• pp.51-52
• /
• 2017

• Journal of Internet of Things and Convergence
• /
• v.6 no.1
• /
• pp.1-8
• /
• 2020

Following the advanced fuel cycle initiative (AFCI) promotions in the United States, the analytic proposition for global fuel cycle initiative (GFCI) has been investigated using dynamical simulations. The political and economic aspects are considered simultaneously due to the particular characteristics of the nuclear materials. The spent nuclear fuels (SNFs) are treated as the reprocessing by the nuclear non-proliferation treaty (NPT) exemption nations and the NPT excluded nations. Otherwise, the pyroprocessing and repository can be done without NPT restriction. In addition, the international trade is considered as the economic aspect where the energy production is a key issue of the GFCI. The dynamical simulations have been done until 2050. The result of the International Trade shows the gradually increasing shape. Additionally, the Nuclear Power Plant Operation shows the increasing by stepwise shape.

• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.675-682
• /
• 2013

The sequential separation process, composed of an oxygen sparging process for separating lanthanides and a zone freezing process for separating Group I and II fission products, was evaluated and tested with a surrogate eutectic waste salt generated from pyroprocessing of used metal nuclear fuel. During the oxygen sparging process, the used lanthanide chlorides (Y, Ce, Pr and Nd) were converted into their sat-insoluble precipitates, over 99.5% at $800^{\circ}C$; however, Group I (Cs) and II (Sr) chlorides were not converted but remained within the eutectic salt bed. In the next process, zone freezing, both precipitation of lanthanide precipitates and concentration of Group I/II elements were preformed. The separation efficiency of Cs and Sr increased with a decrease in the crucible moving speed, and there was little effect of crucible moving speed on the separation efficiency of Cs and Sr in the range of a 3.7 - 4.8 mm/hr. When assuming a 60% eutectic salt reuse rate, over 90% separation efficiency of Cs and Sr is possible, but when increasing the eutectic salt reuse rate to 80%, a separation efficiency of about 82 - 86 % for Cs and Sr was estimated.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.1
• /
• pp.85-89
• /
• 2010

The uranium ingot casting process is one of the steps which consolidate uranium deposits produced by electrorefiner as an ingot form in a pryprocessing technique. This paper introduces new design concept of the ingot casting equipment and the performance test results of the lab-scale ingot casting equipment fabricated based on the design concept. Casting equipment produces the uranium ingot by pouring an uranium melt into a mold by tilting a melting crucible. Also it is equipped with a cup which is able to continuously feed uranium deposits into a melting crucible. The productivity could be significantly enhanced by introducing the continuous operation concept.

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.2
• /
• pp.73-78
• /
• 2009

The decay heat of Cs and Sr contained in a LiCl waste salt, generated from an electrolytic reduction process in pyroprocessing of spent nuclear fuel, has been calculated. The calculation has been carried out under some assumptions that most of the LiCl waste is purified and recycled to main process, and the residual is fabricated to make a waste form. As a result, the decay heat from daughter nuclides such as Ba and Y seems to be maximum 4.6 times higher than that from their parent nuclides such as Cs and Sr. The thermal release from Cs and Sr in the LiCl waste is the maximum around the first one month, so an cooling system operation for some time at the beginning would be suggested to control a rapid increase in the temperature of the LiCl waste salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.15 no.2
• /
• pp.117-124
• /
• 2017

Densities of molten salt mixtures of eutectic LiCl-KCl with $UCl_3$, $CeCl_3$, or $LaCl_3$ at various concentrations (up to 13 wt%) were measured using a liquid surface displacement probe. Linear relationships between the mixture density and the concentration of the added salt were observed. For $LaCl_3$ and $CeCl_3$, the measured densities were significantly higher than those previously reported from Archimedes' method. In the case of $LiCl-KCl-UCl_3$, the data fit the ideal mixture density model very well. For the other salts, the measured densities exceeded the ideal model prediction by about 2%.

• Nuclear Engineering and Technology
• /
• v.46 no.3
• /
• pp.395-406
• /
• 2014

This study provides a fundamental understanding of a cold finger melt crystallization technique by exploring the heat and mass transfer processes of cold finger separation. A series of experiments were performed using a simplified LiCl-CsCl system by varying initial CsCl concentrations (1, 3, 5, and 7.5 wt%), cold finger cooling rates (7.4, 9.8, 12.3, and 14.9 L/min), and separation times (5, 10, 15, and 30 min). Results showed a potential recycling rate of 0.36 g/min with a purity of 0.33 wt% CsCl in LiCl. A CsCl concentrated drip formation was found to decrease crystal purity especially for smaller crystal formations. Dimensionless heat and mass transfer correlations showed that separation production is primarily influenced by convective transfer controlled by cooling gas flow rate, where correlations are more accurate for slower cooling gas flow rates.

• Nuclear Engineering and Technology
• /
• v.52 no.7
• /
• pp.1452-1461
• /
• 2020

A triple bubbler sensor was tested in LiCl-KCl molten salt from 450 to 525 ℃ in a transparent furnace to validate thermal-expansion corrections and provide additional molten salt data sets for calibration and validation of the sensor. In addition to these tests, a model was identified and further developed to accurately determine the density, surface tension, and depth from the measured bubble pressures. A unique feature of the model is that calibration constants can be estimated using independent depth measurements, which allow calibration and validation of the sensor in an electrorefiner where the salt density and surface tension are largely unknown. This model and approach were tested using the current and previous triple bubbler data sets, and results indicate that accuracies are as high as 0.03%, 4.6%, and 0.15% for density, surface tension, and depth, respectively.