• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.199-206
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• 2012

Using fast neutron Generation IV reactors, recycling of americium and curium may become feasible. The detrimental impact of americium on safety parameters has recently been quantified in terms of a power penalty for surviving a given set of transients in sodium fast reactors. In the present paper, a review of the physical reasons for the adverse effect of americium is provided, and different Gen-IV technologies are assessed with respect to their capability of hosting americium in the fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.107-113
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• 2010

In this study, the adsorption reactions in the binary component system such as kaolinite-humic acid, kaolinite-americium and humic acid-americium were investigated. After performing the basic physico-chemical properties of the kaolinite, the adsorption reactions of the humic acid on the kaolinite were carried out with varying concentration of humic acid and ion strength, and pH. With increasing HA concentration and pH, the sorption of HA onto KA decreased, while the sorption of HA onto KA increased with increasing ionic stre ngth. Also, with varying pH, the adsorption reactions of the americium-kaolinite and americium-humic acid were studied. In the acid and neutral region, Am easily adsorbed on the HA, while the sorption of Am on the HA in the alkali region decreased because of electrostatic repulsion. The results from these studies make it possible to understand the characteristics of adsorption behaviour of the americium by the humic acid in the water environment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.399-410
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• 2022

The solubility and species distribution of radionuclides in groundwater are essential data for the safety assessment of deep underground spent nuclear fuel (SNF) disposal systems. Americium is a major radionuclide responsible for the long-term radiotoxicity of SNF. In this study, the solubility of americium compounds was evaluated in synthetic groundwater (SynDB3), simulating groundwater from the DB3 site of the KAERI Underground Research Tunnel. Geochemical modeling was performed using the ThermoChimie_11a thermochemical database. Concentration of dissolved Am(III) in Syn-DB3 in the pH range of 6.4-10.5 was experimentally measured under over-saturation conditions by liquid scintillation counting over 70 d. The absorption spectra recorded for the same period suggest that Am(III) colloidal particles formed initially followed by rapid precipitation within 2 d. In the pH range of 7.5-10.5, the concentration of dissolved Am(III) converged to approximately 2×10^-7 M over 70 d, which is comparable to that of the amorphous AmCO₃OH(am) according to the modeling results. As the samples were aged for 70 d, a slow equilibrium process occurred between the solid and solution phases. There was no indication of transformation of the amorphous phase into the crystalline phase during the observation period.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.787-791
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• 2017

Americium-beryllium (Am-Be; n, ${\gamma}$) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2454-2463
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• 2021

The objective of this research is to the use of americium (AmO₂) as a burnable absorber effectively instead of conventional gadolinium (Gd₂O₃) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO₂ as a burnable absorber. From these studies, it is found that AmO₂ can decrease the excess reactivity like Gd₂O₃ without changing the criticality life span and enrichment of ²³⁵U. A homogeneous mixture of the 0.20% AmO₂+ 4.95% enriched UO₂ fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO₂+UO₂ in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.594-595
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.10a
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• pp.363-364
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• 2011

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.9-20
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• 2008

Since plutonium accounts for 40-50% of the power produced by uranium fuels, spent fuel contains only residual plutonium. Management of this plutonium is one of the aspects influencing the choice of a fuel cycle back-end option: reprocessing, direct disposal or wait-and-see. Different grades and qualities of plutonium exist depending from their specific generation conditions; all are valuable fissile material. Safeguard authorities watch the inventories of civil plutonium, but access to those data is restricted. Independent evaluations have led to an estimated current inventory of 220t plutonium in total (spent fuel, separated civil plutonium and military plutonium). If used as MOX fuel, it would be sufficient to feed all the PWRs and BWRs worldwide during 7 years or to deploy a FBR park corresponding to 150% of today' s installed nuclear capacity worldwide, which could then be exploited for centuries with the current stockpile of depleted and spent uranium. The energy potential of plutonium deteriorates with storage time of spent fuel and of separated plutonium, due to the decay of $^{241}Pu$, the best fissile isotope, into americium, a neutron absorber. The loss of fissile value of plutonium is more pronounced for usage in LWRs than in FBR. However, keeping the current plutonium inventory for an expected future deployment of FBRs is counterproductive. Recycling plutonium reduce the required volume for final disposal in an underground repository and the cost of final disposal. However, the benefits of utilizing an energy resource and of reducing final disposal liabilities are not the only aspects that determine the choice of a back-end policy.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1277-1283
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• 2021

It is important for operators of poly-pipelines in petroleum industry to continuously monitor characteristics of transferred fluid such as its type and amount. To achieve this aim, in this study a dual energy gamma attenuation technique in combination with artificial neural network (ANN) is proposed to simultaneously determine type and amount of four different petroleum by-products. The detection system is composed of a dual energy gamma source, including americium-241 and barium-133 radioisotopes, and one 2.54 cm × 2.54 cm sodium iodide detector for recording the transmitted photons. Two signals recorded in transmission detector, namely the counts under photo peak of Americium-241 with energy of 59.5 keV and the counts under photo peak of Barium-133 with energy of 356 keV, were applied to the ANN as the two inputs and volume percentages of petroleum by-products were assigned as the outputs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.397-410
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• 2018

When considering the long-term safety assessment of spent-nuclear fuel management, americium is one of the most radio-toxic actinides. Although spectroscopic methods are widely used for the study of actinide chemistry, application of those methods to americium chemistry has been limited. Herein, we purified $^{241}Am$ to obtain a highly pure stock solution required for spectroscopic studies. Quantitative and qualitative analyses of purified $^{241}Am$ were carried out using liquid scintillation counting, and gamma and alpha radiation spectrometry. Highly sensitive absorption spectrometry coupled with a liquid waveguide capillary cell and time-resolved laser fluorescence spectroscopy were employed for the study of Am(III) hydrolysis and oxalate (Ox) complexation. $Am^{3+}$ ions under acidic conditions exhibit maximum absorbance at 503 nm, with a molar absorption coefficient of $424{\pm}8cm^{-1}{\cdot}M^{-1}$. $Am(OH)_3(s)$ colloidal particles formed under near neutral pH conditions were identified by monitoring the absorbance at around 506-507 nm. The formation of ${Am(Ox)_3}^{3-}$ was detected by red-shifts of the absorption and luminescence spectra of 4 and 5 nm, respectively. In addition, considerable enhancements of the luminescence intensities were observed. The luminescence lifetime of ${Am(Ox)_3}^{3-}$ increased from 23 to 56 ns, which indicates that approximately six water molecules are replaced by carboxylate ligands in the inner-sphere of the Am(III). These results suggest that ${Am(Ox)_3}^{3-}$ is formed through the bidentate coordination of the oxalate ligands.