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At least 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Equivalent Fundamental-Mode Source Simulations for Spherical Uranium and Plutonium Systems

It is important in neutron noise measurements and reactor startup to understand how the source distribution in a system compares to the equivalent fundamental-mode (EFM) source. Deviation from the fundamental mode source distribution increases as the system $\kappa$ eff increases. A correction factor, g *, is a measure of this deviation, and is the focus of this work. Two methods were utilized to simulate g *. These methods are compared to each other and (where applicable) previous results. The effects associated with nuclear material inner radius, materials of reflectors, and reflector thickness are also investigated.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Distillation of Americium from Plutonium Americium Oxide

Understanding the changes that fuel undergoes inside a reactor and understanding the properties of recycled fuel to provide a foundation for the development of a sustainable nuclear fuel cycles are two of the main objectives of Fuel Cycle Research and Development (FCRD). Research on transmutation fuel aims to develop technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby decreasing both the volume of material requiring disposal and the long-term radiotoxicity of the waste. Since there is no recycled nuclear fuel available for study within the United States, studies of this type of fuel must be conducted on fabricated fuel. Americium is an important component of recycled fuel and, thus, is a vital additive in the fabrication process. Unfortunately, the supplies of pure americium metal are dwindling. Idaho National laboratory (INL) has acquired a stockpile of americium- bearing material from around the DOE complex which can be purified to supply the needs of the fuel fabrication process. In order to make this material usable, it must be distilled (or, in some cases, reduced from an oxide) to a material pure enough to be used in the fuel fabrication process. The americium distillation experiments were set up to develop a process to convert the available americium bearing material into pure americium metal. This work gives important insight into the best parameters to use in future americium distillation runs. Since future starting material is likely to contain significantly more oxide, it will be necessary to add a reducing agent such as lanthanum in order to reduce the oxide material prior to distillation. Early experiments show that the addition of lanthanum does not significantly change the purity of the end product. Soak time was found to be beneficial in producing a higher quantity of pure material. All experiments show that the material preferentially deposits on the walls of the crucible rather than on the cold finger as was originally intended. Future experiments will aim to deposit material on the cold finger rather than the crucible walls in order to make material retrieval easier as well as to increase yield. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Colloidal Behavior of Plutonium Oxide in Concentrated Electrolyte Solutions

The Hanford Site in Washington State manages legacy high-level radioactive waste streams that display major chemistry and engineering challenges, including the high salt levels and pH values that correspond to conditions under which many classical concepts describing chemical reactivity cannot be applied. One particular challenge that needs to be tackled at Hanford is that Pu concentrations, [Pu], in the soluble phases of the tank wastes are higher than expected based on the solubility of crystalline PuO2, which is widely accepted to be caused by the formation of PuO2 colloid, consisting of nano- to submicron-sized particles (PuO2 NPs). Fundamental research underpinning the behavior of PuO2 NPs under conditions not only relevant to the Hanford tank waste but at high ionic strength in general is needed to reliably predict the chemical reactivity of PuO2 NPs and develop engineering solutions to safely and efficiently process high-level radioactive waste into forms suitable for long-term storage. In this work, we study the behavior of PuO2 NPs (particle size ~100 nm) under high ionic strength conditions by reacting it with highly concentrated (up to 5 M) salt solutions. We explore different electrolyte compositions to elucidate the impact of different anions (NO3-, Cl-, ClO4-, SO42-, C2O42-, CO32-) on the stability of PuO2 NP in the acidic and alkaline pH regime. PuO2 NP aggregation and precipitation as function electrolyte concentration is tracked by a combination of liquid scintillation counting, dynamic light scattering for determination of particle size distributions, and zeta potentials as a proxy for particle charge. At acidic pH, electrolytes containing non-coordinating anions, such as NaNO3, NaCl, and NaClO4 mostly stabilize PuO2 NPs over a large electrolyte concentration range, showing only subtle differences in their reactivity. Other electrolyte anions show a more pronounced effect on the PuO2 NP stability: SO42-, binds directly to the particles’ surface, reverses the particle charge, and precipitates the PuO2 NPs efficiently even at intermediate sulfate concentrations (>0.1 M). In contrast, C2O42- is found to lead to high [Pu] in solution, in the milli-molar range, even at mildly acidic pH (~4). Thermodynamic modeling of the dissolved Pu concentrations using PHREEQC is unable to predict the observed [Pu] in the acidic pH regime, supporting the influence of colloids in maintaining elevated [Pu]. It is noteworthy that the current thermodynamic databases do not include constants for colloidal Pu phases and cannot accurately predict many of the high ionic strength solutions relevant to this work. The mechanisms and models responsible for these observations will need further investigation in the future. At high pH values (~12), PuO2 NPs exhibits classical sol-gel chemistry, meaning that upon destabilization of the colloidal sol, for example by addition of concentrated NaOH, highly porous and viscid PuO2 coagulates are formed that consist of a three-dimensional network likely held together by physical interactions. The PuO2 NP coagulate shows no significant reversibility of the aggregation when contacted with concentrated brines; however, PuO2 NPs can be efficiently resuspended in solution by addition of diluted electrolytes, alkaline solutions containing high amounts of carbonate, or simple addition of water. Especially carbonate is shown to stabilize PuO2 NPs in solution at high pH, characterized by stable colloidal suspensions that are resistant against sedimentation during centrifugation. Thermodynamic modeling of the carbonate system was able to predict an increasing dissolved Pu concentration with increasing carbonate concentration. However, the model was profoundly sensitive to the fixed redox potential and does not include any thermodynamic constants for colloidal Pu species.

Neumann, Julia↗