L.M. Toth

Effect of curing temperature on green light emission from Er3+-doped sol-gel silica glass

Abstract Green and less intense red emissions from Er3+ (4f11)-doped sol-gel silica samples (prepared from tetraethylorthosilicate at room temperature and dried at 70°C) under 488 nm excitation were monitored as a function of sample curing temperature. These green and red emission bands, centered at 550 and 660 nm, are the 4 S 3 2 → 4 I 15 2 and 4 F 9 2 → 4 I 15 2 transitions in Er3+ ions, respectively. The emission intensities were found to increase with increasing sample curing temperature ranging from 200 to 1000°C. Green upconversion emission from Er3+ ions in the same samples was also recorded under 647.1 nm light excitation. Its efficiency was enhanced with increased sample curing temperature. This enhancement is attributed to the Er3+-coordination environment produced by different sample curing temperatures. Increasing curing temperatures produces increasing degrees of dehydration and de-alcoholation around the Er3+ ions, thus resulting in a decrease in the non-radiative emission rate.

Citrate-Based “TALSPEAK” Actindide-Lanthandide Separation Process

Abstract Lanthanide elements are produced in relatively high yield by fission of235 U. Almost all the lanthanide isotopes decay to stable nonradioactive lanthanide isotopes in a relatively short time. Consequently, it is highly advantageous to separate the relatively small actinide fraction from the relatively large quantities of lanthanide isotopes. The TALSPEAK process (Trivalent Actinide Lanthanide Separations by Phosphorusreagent Extraction from Aqueous Complexes) is one of the few means available to separate the trivalent actinides from the lanthanides. Previous work based on the use of lactic or glycolic acid has shown deleterious effects of some impurity ions such as zirconium (IV), even at concentrations on the order of 10−4 M. Other perceived problems were the need to maintain the pH and reagent concentrations within a narrow range and a significant solubility of the organic phase at high carboxylic acid concentrations. Our cold experiments showed that replacing the traditional extractants glycol...

Molten fluoride fuel salt chemistry

The chemistry of molten fluorides is traced from their development as fuels in the Molten Salt Reactor Experiment with important factors in their selection being discussed. Key chemical characteristics such as solubility, redox behavior, and chemical activity are explained as they relate to the behavior of molten fluoride fuel systems. Development requirements for fitting the current state of the chemistry to modern nuclear fuel system are described. It is concluded that while much is known about molten fluoride behavior which can be used effectively to reduce the amount of development required for future systems, some significant molten salt chemical questions must still be addressed.

Fluorine generation by gamma radiolysis of a fluoride salt mixture

Abstract Fluoride salt mixtures of composition, LiF–BeF2–ZrF4–UF4 (64.5, 30.3, 5.0, 0.13 mol%) were irradiated using 107–108 R/h gamma radiation to achieve high doses. Under these conditions, F2 is released from the salt and accumulates as an F2 overpressure until a 2 mol% damage limit in the crystal is reached. At this point, the rate of F2 generation is equivalent to the rate of recombination with active metal sites accompanying the radiolytic damage in the salt. The activation energy for the rate of recombination has been found to be 39 kJ/mol; and for the conditions of irradiation used here, no F2 should be released above 150°C.

Citrate based ``TALSPEAK`` lanthanide-actinide separation process

The potential hazard posed to future generations by long-lived radionuclides such as the transuranic elements (TRU) is perceived as a major problem associated with the use of nuclear power. TRU wastes have to remain isolated from the environment for ``geological`` periods of time. The costs of building, maintaining, and operating a ``geological TRU repository`` can be very high. Therefore, there are significant economical advantages in segregating the relatively low volume of TRU wastes from other nuclear wastes. The chemical behavior of lanthanides and actinides, 4f and 5f elements respectively, is rather similar. As a consequence, the separation of these two groups is difficult. The ``TALSPEAK`` process (Trivalent Actinide Lanthanide Separations by Phosphorus-reagent Extraction from Aqueous Complexes) is one of the few means available to separate the trivalent actinides from the lanthanides. The method is based on the preferential complexation of the trivalent actinides by an aminopolyacetic acid. Cold experiments showed that by using citric acid the deleterious effects produced by impurities such as zirconium are greatly reduced.

Gamma radiolysis studies of uranyl fluoride

Summary The safe handling and storage of radioactive materials require an understanding of the effects of radiolysis on those materials. Radiolysis may result in the production of gases (e.g., corrosives) or pressures that are deleterious to storage containers. A study has been performed to address these concerns as they relate to the radiolysis of residual fluoride compounds in uranium oxides. Samples of UO2F2·xH2O and U3O8 (with ∼1.4 wt.% fluorine content) were irradiated in a 60Co source and in spent nuclear fuel (SNF) elements from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. Container pressures were monitored throughout the irradiations, and gas and solid samples were analyzed after the irradiations. The irradiation of UO2F2·xH2O produced O2 – with G(O2)-values ranging from 0.007 to 0.03 molecules of O2 produced per 100 eV. Neither F2 nor HF was produced by the irradiations. Chemical analyses of solid samples showed that some of the uranium was reduced from U(VI) to U(IV). A saturation damage limit for the UO2F2·xH2O was demonstrated by using the HFIR SNF elements, and the limit was found to be 7–9% at ∼108 rad/h). It is shown that the covalently bonded oxygen is more susceptible to radiation damage than is the ionically bonded fluorine. Irradiation of U3O8 (with ∼1.4 wt.% fluorine content) resulted in neither gas production nor a pressure increase. These experiments led to the conclusion that during long-term storage U3O8 is safe from overpressurization and the production of corrosives caused by gamma radiolysis of residual fluorides.

Some investigations of the reaction of activated charcoal with fluorine and uranium hexafluoride

Abstract Since the 1969 shutdown of the Molten Salt Reactor Experiment (MSRE) at Oak Ridge, radiolytically generated fluorine (F 2 ) and 233-uranium hexafluoride ( 233 UF 6 ) migrated from the fuel storage tanks through gas piping to a charcoal bed. This report addresses the carbon–fluorine–uranium chemistry under conditions reproducing those found in the charcoal beds. Laboratory analysis of the reaction products has been extensive and includes electron spectroscopy for chemical analysis (ESCA), nuclear magnetic resonance (NMR), FTIR and Raman spectroscopy, TGA-DTA as well as a host of other techniques. The chemical identity, stoichiometry, structure, thermochemistry, and potential for energetic decomposition of the primary reaction product, ‘fluorinated charcoal’, was determined. As a result of this research, remedial solutions were developed and implemented.

The Influence of Lewis Acid/Base Chemistry on the Removal of Gallium by Volatility from Weapons-Grade Plutonium Dissolved in Molten Chlorides

Abstract It has been proposed that GaCl3 can be removed by direct volatilization from a Pu-Ga alloy that is dissolved in a molten chloride salt. Although pure GaCl3 is quite volatile (boiling point: 201°C), the behavior of GaCl3 dissolved in chloride salts is quite different because of solution effects and is critically dependent upon the composition of the solvent salt (i.e., its Lewis acid/base character). In this technical note, the behavior of gallium in prototypical Lewis acid and Lewis base salts is contrasted. It is found that gallium volatility is suppressed in basic melts and promoted in acidic melts. These results have an important influence on the potential for simple gallium removal in molten salt systems.

Evaluation of Process That Might Lead to Separation of Actinides in Waste Storage Tanks Under Alkaline Conditions

This study addresses the physical-chemical processes that might naturally or inadvertently occur and that would lead to a separation of the poisoning nonfissionable actinides (232Th, 238U) from the fissionable ones (239Pu, 235U) by selective dissolution and redeposition over a prolonged storage of the waste. Of the various chemistries that were evaluated, carbonate complexation reaction is the most plausible means of achieving the separation of these actinides. Carbonate ions (formed by the dissolution and hydrolysis of atmospheric CO2) can selectively dissolve the actinide oxides through the formation of soluble carbonate complexes, which could result in the separation of poisoning actinides from the fissionable ones. The concentrations of these soluble carbonate species are dependent on the pH, temperature, and other ions; therefore, changes in any of these parameters over time—especially cyclic changes (daily or seasonal)—could cause a selective dissolution and redeposition of the more soluble species away from the less soluble ones. Detailed calculations using the stability constants for the carbonates have shown that the most likely pH range for this process to occur is pH = 10–11. Increased solubility through reaction with organic complexants such as EDTA was also considered, and while it presents a situation similar to carbonate complexation and similar potential for autoseparation of the actinides in the waste tanks, it would require the uncontrolled dumping of large amounts of complexants into the storage tanks.

A Simple Kinetic Model for the Alpha Radiolysis of Water Sorbed on NpO2

Abstract Alpha radiolysis experiments have been performed on NpO2 that contains sorbed moisture. A high dose rate to the sample was achieved by spiking it with ~7000 ppm 244Cm during preparation. Pressure monitoring of sample containers showed that a low, steady-state pressure plateau is reached. This plateau indicates a situation in which the forward reaction (i.e., radiolysis of water) is equal to the back reaction (i.e., the reformation of H2O). In this technical note, a simple kinetic model that can be used for predicting steady-state pressures under practical conditions is described.