• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1145-1152
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• 2024

To realize an initial objective of the negative ion-based neutral beam injection (N-NBI) at the Comprehensive Research Facility for Fusion Technology (CRAFT) test facility, which targets an H⁰ beam power of 2 MW at an energy of 200-400 keV and a pulse duration of 100 s, it is crucial to study the cesium dynamics of the negative ion source. Here a numerical simulation program CSFC3D is developed and applied to simulate the distribution and time dynamics of cesium during short pulses. The calculations show that most of the cesium on the plasma grid (PG) area originates from the release of cesium that is accumulated within the ion source in the plasma phase. Increasing the wall temperature reduces the loss of cesium on the wall of the ion source. Furthermore, the thickness of the cesium monolayer is directly influenced by the PG temperature. Both simulated and experimental results demonstrate that maintaining the PG temperature between 180 ℃ and 200 ℃ is essential for enhancing the performance of the ion source and optimizing the cesium behavior.

• Journal of Korean Society of Water and Wastewater
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• v.27 no.2
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• pp.155-164
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• 2013

The main purpose of this research was to examine the adsorption/ion exchange characteristics of radioactive species such as cesium and strontium onto synthetic Na-birnessite (sodium-birnessite). As part of efforts to investigate the sorption behavior of cesium and strontium onto synthetic Na-birnessite, batch isotherm tests were performed under different experimental conditions. Na-birnessite was synthesized by the oxidation of $Mn^{2+}$ ions in sodium hydroxide solution. The synthetic Na-birnessite was characterized by powder x-ray diffraction (XRD), scanning electron microscopy (SEM), energy-dispersive x-ray spectroscopy (EDS), and Brunauer-Emmett-Teller (BET) surface area analysis. Cesium and strontium concentrations were determined by atomic absorption spectroscopy (AAS). The removal efficiency of strontium by Na-birnessite was around 95 % which was much higher than that of cesium (~ 32 %). The results imply that strontium has a higher affinity for Na-birnessite than cesium because strontium, divalent cation leads to larger electrostatic attraction than monovalent cesium.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.167-173
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• 2004

The amount of cesium released from the leaching of spent fuels in contact with and without the compacted bentonite bloc]t which was compacted as the density of $1.4g/\textrm{cm}^3$, up to 5.7 years were measured and the empirical formula of the fractional release rate of cesium were derived from these measured values. The empirical formulas show that the long-term release rate of cesium under a repository would become a constant, as about $3{\times}10_{-6}$ fraction/day, after a certain period. The cumulative fractions of cesium released from the spent fuel with bentonite and with copper and stainless steel sheets were steadily increased, but the fraction from bare fuel was rapidly increased and then sluggishly increased. However, the remained value except its gap inventory from the cumulative fraction of cesium released from bare fuel was almost very close to the others. This suggests that the initial release of cesium from bare fuel might be dependant on its gap inventory.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.80-89
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• 2002

A laboratory investigation of the behavior of radioactive metals under the various waste combustion atmospheres was conducted to predict the parameters that influence their partitioning behavior during waste incineration. Neodymium, samarium, cerium, gadolinium, cesium and cobalt were used as non-radioactive surrogate metals that are representative of uranium, plutonium, americium, curium, radioactive cesium, and radioactive cobalt, respectively. Except for cesium, all of the investigated surrogate metal compounds converted into each of their stable oxides at medium temperatures from 400 to 90$0^{\circ}C$, under oxygen- deficient and oxygen-sufficient atmospheres (0.001-atm and 0.21-atm $O_2$). At high temperatures above 1,40$0^{\circ}C$, cerium, neodymium and samarium in the form of their oxides started to vaporize but the vaporization rates were very slow up to 150$0^{\circ}C$ . Inorganic chlorine (NaCl) as well as organic chlorine (PVC) did not impact the volatility of investigated Nd$_2$O$_3$, CoO and Cs$_2$O. The results of laboratory investigations suggested that the combustion chamber operating parameters affecting the entrainment of particulate and filtration equipment operating parameters affecting particle collection efficiency be the governing parameters of alpha radionuclides partitioning during waste incineration.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.350-353
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• 2016

Background: Various characteristics such as size distribution, chemical component and radio-activity have been analyzed for radioactive aerosols released from Fukushima Daiichi Nuclear Power Plant. Measured results for radioactive aerosols suggest that the potential transport medium for radioactive cesium was non-sea-salt sulfate. This result indicates that cesium isotopes would preferentially attach with sulfate compounds. In the present work the attachment behavior of fission products to aqueous solution aerosols of sodium salts has been studied using a generation system of solution aerosols and spontaneous fission source of $^{248}Cm$. Materials and Methods: Attachment ratios of fission products to the solution aerosols were compared among the aerosols generated by different solutions of sodium salt. Results and Discussion: A significant difference according as a solute of solution aerosols was found in the attachment behavior. Conclusion: The present results suggest the existence of chemical effects in the attachment behavior of fission products to solution aerosols.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.409-414
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• 1996

Radionuclides such as Cs and Sr were removed from dilute aqueous solutions by means of inorganic adsorbents, 13X and chabazite. The physical adsorption obeyed the DA equation and non-equilibrium dynamic adsorption model, which describes surface diffusion mechanism with the DA equation, simulated the adsorption behavior of cesium and strontium on zeolite in fixed bed adsorbers. The dynamic model simulated the adsorption behavior of cesium and strontium.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.125-137
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• 2019

The purpose of this paper is to study the source term behavior after severe accidents by using a semi-kinetic model for simulation and calculation of in-containment activity. The reactor containment specification and the safety features of the containment under different accident conditions play a great role in evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of in-containment isotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements of iodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release under LOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetric leakage rate for noble gasses is important during the accident, while the influence of deposition on the containment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution in removal of activity during evolution of the accident.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.859-866
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• 2015

The aim of this work is to investigate the transport mechanism of radioactive nuclides through the reverse osmosis (RO) membrane and to estimate its effectiveness for nuclide separation from radioactive liquid waste. An analytical model is developed to simulate the RO separation, and a series of experiments are set up to confirm its estimated separation behavior. The model is based on the extended Nernst-Plank equation, which handles the convective flux, diffusive flux, and electromigration flux under electroneutrality and zero electric current conditions. The distribution coefficient which arises due to ion interactions with the membrane material and the electric potential jump at the membrane interface are included as boundary conditions in solving the equation. A high Peclet approximation is adopted to simplify the calculation, but the effect of concentration polarization is included for a more accurate prediction of separation. Cobalt and cesium are specifically selected for the experiments in order to check the separation mechanism from liquid waste composed of various radioactive nuclides and nonradioactive substances, and the results are compared with the estimated cobalt and cesium rejections of the RO membrane using the model. Experimental and calculated results are shown to be in excellent agreement. The proposed model will be very useful for the prediction of separation behavior of various radioactive nuclides by the RO membrane.

• Bulletin of the Korean Chemical Society
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• v.31 no.10
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• pp.2867-2872
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• 2010

Two new energetic organic alkali metal salts, 1,1-diamino-2,2-dinitroethylene rubidium salt [Rb(FOX-7)${\cdot}H_2O$] and 1,1-diamino-2,2-dinitroethylene cesium salt [Cs(FOX-7)${\cdot}H_2O$], were synthesized by reacting of 1,1-diamino-2,2-dinitroethylene (FOX-7) and rubidium chloride or cesium chloride in alkali methanol aqueous solution, respectively. The thermal behaviors of Rb(FOX-7)${\cdot}H_2O$ and Cs(FOX-7)${\cdot}H_2O$ were studied with DSC and TG methods. The critical temperatures of thermal explosion of the two compounds are 216.22 and $223.73^{\circ}C$, respectively. Specific heat capacities of the two compounds were determined with a micro-DSC method, and the molar heat capacities are 217.46 and $199.47\;J\;mol^{-1}\;K^{-1}$ at 298.15 K, respectively. The adiabatic times-to-explosion were also calculated to be a certain value of 5.81 - 6.36 s for Rb(FOX-7)${\cdot}H_2O$, and 9.92 - 10.54 s for Cs(FOX-7)${\cdot}H_2O$. After FOX-7 becoming alkali metal salts, thermal decomposition temperatures of the compounds heighten with the rise of element period, but thermal decomposition processes become intense.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2878-2887
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• 2021

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1-4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.