PuF3 Research Articles

Stabilization of Plutonium(V) Within a Crown Ether Inclusion Complex

Crystalline coordination complexes of actinides, especially in atypical oxidation states, are not only fundamentally important for expanding the notably limited knowledge on the bonding nature of a...

Investigation of PuF3 and AmF3 Solubility in 73LiF–27BeF2 Melt

Using an experimental stand for studying the solubility of fluorides, measurements were performed at 550–700°C with step 50°C and the solubility of plutonium and americium trifluorides in melt with molar composition 73LiF–27BeF2 was calculated. Previously, such measurements were not performed for plutonium and americium trifluoride. The temperature dependence of the solubility of plutonium and americium trifluoride for the given salt system was constructed. The obtained solubility for plutonium trifluoride is appreciably higher than in a similar salt system with a higher content of beryllium difluoride. The obtained data on the solubility of americium and plutonium trifluoride can be used to select the minimum temperature of the fuel salt of a molten-salt reactor.

Vaporization behaviour of a PuF3-containing fuel mixture for the Molten Salt Fast Reactor

The mixture LiF-ThF4-UF4-PuF3 (77.5–6.6-12.3–3.6 mol%), a fuel option for the Molten Salt Fast Reactor (MSFR), has been measured by differential scanning calorimetry (DSC) for determination of phase transitions, and by Knudsen effusion mass spectrometry (KEMS) for measurement of partial vapour pressures. The boiling point of the mixture was determined by extrapolation of total vapour pressure to 1 bar. Thermodynamic calculations were performed and compared with the experimental results. Novel experimental data on pure plutonium trifluoride are presented: melting point, vaporization enthalpy, vapour pressure and ionization energies by electron impact.

Synthesis of plutonium trifluoride by hydro-fluorination and novel thermodynamic data for the PuF3-LiF system

PuF3 was synthetized by hydro-fluorination of PuO2 and subsequent reduction of the product by hydrogenation. The obtained PuF3 was analysed by X-Ray Diffraction (XRD) and found phase-pure. High purity was also confirmed by the melting point analysis using Differential Scanning Calorimetry (DSC). PuF3 was then used for thermodynamic assessment of the PuF3-LiF system. Phase equilibrium points and enthalpy of fusion of the eutectic composition were measured by DSC. XRD analyses of selected samples after DSC measurement confirm that after solidification from the liquid, the system returns to a mixture of LiF and PuF3.

Application of Vacuum Salt Distillation Technology for the Removal of Fluoride and Chloride from Legacy Fissile Materials

Between September 2009 and January 2011, the Savannah River National Laboratory (SRNL) and the Savannah River Site (SRS) HB-Line Facility designed, developed, tested, and successfully deployed a production-scale system for the distillation of sodium chloride (NaCl) and potassium chloride (KCl) from plutonium oxide (PuO2) scrap. Subsequent efforts adapted the vacuum salt distillation (VSD) technology for the removal of chloride and fluoride from less-volatile halide salts at the same process temperature and vacuum. Calcium chloride (CaCl2), calcium fluoride (CaF2), and plutonium trifluoride (PuF3) were of particular concern. To enable the use of the same operating conditions for the distillation process, SRNL employed in situ exchange reactions with alkali metal hydroxides (e.g., potassium hydroxide) to convert the less-volatile halide salts to volatile compounds that facilitated the distillation of halide without removal of plutonium. SRNL demonstrated the removal of halide from CaCl2, CaF2, and PuF3 below 1000°C using VSD technology.

Compatibility of Selected Ni-Based Alloys in Molten Li,Na,Be/F Salts with PuF3 and Tellurium Additions

In the last years, there has been a real growth of interest in the use of high-temperature molten salt fluorides as coolants and fuel salts in nuclear power systems. For all molten salt reactor (MSR) concepts, material selection is a very important issue. This paper summarizes results of recent work done within the ISTC#1606 project and the present state of knowledge about container materials for MSRs. The central focus is the compatibility of Ni-based alloys with the molten Li,Na,Be/F salt system as applied to the primary circuit of the MOlten Salt Actinide Recycler & Transmuter (MOSART) fueled with different compositions of actinide trifluorides from light water reactor spent fuel without U-Th support. Results from recent studies with a Li,Na,Be/F thermal convection loop at temperatures up to 700°C are presented. Material specimens of three modified Hastelloy N alloys, particularly HN80M-VI with 1% of Nb, HN80MTY with 1% of Al, and MONICR with 2% of Fe, were used for our study in corrosion facilities. Methods to purify the molten salt composition and to improve Ni-based container alloy compatibility by maintaining the salt at a low redox potential are discussed. The effect on materials compatibility of adding plutonium trifluoride and tellurium to the Li,Na,Be/F solvent system is also considered. Last, testing of advanced Ni-based alloys with various compositions to enhance first of all its resistance to tellurium intergranular cracking should be continued in thermal convection loops with a long exposure time for the MOSART fuel salts as well as the novel nonmoderated thorium sustainable MSR concept in the framework of the new ISTC#3749 project.

First-principles calculations of PuO(2+/-x).

The electronic structure of PuO(2+/-x) was studied using first-principles quantum mechanics, realized with the self-interaction corrected local spin density method. In the stoichiometric PuO2 compound, Pu occurs in the Pu(IV) oxidation state, corresponding to a localized f4 shell. If oxygen is introduced onto the octahedral interstitial site, the nearby Pu atoms turn into Pu(V) (f3) by transferring electrons to the oxygen. Oxygen vacancies cause Pu(III) (f5) to form by taking up electrons released by oxygen. At T = 0, the PuO2 compound is stable with respect to free oxygen, but the delicate energy balance suggests the possible deterioration of the material during long-term storage.

Pyrochemical Reprocessing of Weapons Plutonium as Nuclear Fuel for Power Reactors

Experimental work on transferring compact weapons plutonium by pyrochemical methods into its compounds was generalized. Powdered plutonium dioxide was obtained by interaction with moist air at 480–500°C. PuH3 was synthesized by interaction of the metal with hydrogen at 280°C. PuF3 was formed in an interaction of dioxides with HF and H2. The reaction of plutonium hydride with HF also resulted in plutonium trifluoride. PuF4 was synthesized by interaction of PuO2 with HF. In summary, compact plutonium can be transfered by water-free methods into its powdered compounds, which can be used in investigations for developing new types of fuel elements for power reactors. 6 figures, 2 tables, 16 references.

Synthesis of plutonium trifluoride from weapons plutonium as a potential fuel for power reactors

Synthesis of plutonium trifluoride from weapons plutonium as a potential fuel for power reactors

Plutonium Trifluoride as a Fuel for Molten Salt Reactors-Solubility Studies

The feasibility of using plutonium as a fuel in a molten salt reactor has been evaluated by the determination of the solubility of PuF$sub 3$ in a number of salt compositions containing LiF, BeF$sub 2$, and ThF$sub 4$ in the temperature range of interest for such a reactor. An empirical correlation for the solubility of PuF$sub 3$ as a function of salt composition has been obtained from the solubility data to calculate the solubility of PuF$sub 3$ in other salt compositions of interest. The data indicate that PuF$sub 3$ has adequate solubility in salt compositions of interest for a molten salt reactor of either the converter or the breeder type. (auth)

Effect of F − complexing agents on PuF 4 dissolution

Effect of F − complexing agents on PuF 4 dissolution

Optical Emission Spectrographic Determination of Tantalum and Tungsten in Plutonium

A spectrochemical method for the determination of 10–1000 ppm tantalum and 20–1000 ppm tungsten in plutonium metal is described. Chemical separation by precipitation of the plutonium as insoluble plutonium trifluoride, centrifuging, and sampling the supernatant solution is well suited to the glove box operation necessary in handling plutonium. Excitation is by graphite spark; spectra are recorded photographically; and concentrations are determined by standard photometry practices using an emulsion calibration-analytical curve technique with a molybdenum internal standard.

Precipitation of Plutonium Trifluoride

ADVERTISEMENT RETURN TO ISSUEPREVArticleNEXTPrecipitation of Plutonium TrifluorideG. A. Burney and F. W. ToberCite this: Ind. Eng. Chem. Process Des. Dev. 1965, 4, 1, 28–32Publication Date (Print):January 1, 1965Publication History Published online1 May 2002Published inissue 1 January 1965https://doi.org/10.1021/i260013a009RIGHTS & PERMISSIONSArticle Views129Altmetric-Citations6LEARN ABOUT THESE METRICSArticle Views are the COUNTER-compliant sum of full text article downloads since November 2008 (both PDF and HTML) across all institutions and individuals. These metrics are regularly updated to reflect usage leading up to the last few days.Citations are the number of other articles citing this article, calculated by Crossref and updated daily. Find more information about Crossref citation counts.The Altmetric Attention Score is a quantitative measure of the attention that a research article has received online. Clicking on the donut icon will load a page at altmetric.com with additional details about the score and the social media presence for the given article. Find more information on the Altmetric Attention Score and how the score is calculated. Share Add toView InAdd Full Text with ReferenceAdd Description ExportRISCitationCitation and abstractCitation and referencesMore Options Share onFacebookTwitterWechatLinked InReddit PDF (1 MB) Get e-Alerts

Plutonium Metal from Trifluoride

Plutonium trifluoride can be precipitated from a plutonium(III) nitrate solution, dried, and reduced to metal using calcium with an iodine booster. This system, with variations, has been in routine use at the Savannah River Plant for several years to produce 2000-gram buttons of plutonium metal. Criteria that determine crystal size and filtrate loss during precipitation in a two-stage unit appear applicable to the design and operation of multistage units for the precipitation of other compounds. The low heat of reduction of the trifluoride requires 0.6 mole of iodine per mole of plutonium as a booster for good yields. Problems caused by iodine are eliminated by oxidizing anhydrous PuFg with dry oxygen to a mixture of PuF/su 4/ and PuO/sub 2/, which has sufficient heat of reduction to be reduced with calcium alone, in 96% yield. General relationships between plutonium fluorides and oxides, the role of hydrated tetrafluorides, and filtrate recovery are discussed. (auth)

475. The solubility of plutonium trifluoride, plutonium tetrafluoride, and plutonium(IV) oxalate in nitric acid mixtures

The solubility of plutonium trifluoride was measured in mixtures of nitric and hydrofluoric acids of various molarities and tetrafluoride in nitric acid. The effect of fluoride ion on the solubility is discussed and an attempt made to estimate the solubility product of each compound. On the basis of available data, the solubility products appear to be trifluoride 2.5 x 10/sup -16/ and tetrafluoride 6 x 10/sup -20/. The solubility of plutonium(IV) oxalate was determined in mixtures of nitric acid up to 3.52M and oxalic acid up to 0.6M. The minimum solubility occurs in 1M nitric acid to O.O1M oxalic acid, but there is evidence that after 2 to 3 weeks' eqailibration in this mixture an abnormal effect leads to anoma-. lous solubilities. (auth)

Experimental Estimate of the Free Energy of Formation of Plutonium Trifluoride

A study involving equilibration at 1573°K of fused uranium tetrafluoride and molten irradiated uranium has permitted calculation of the free energy of formation of plutonium trifluoride at 1573°K. During equilibration it is assumed that plutonium is extracted into molten uranium fluoride as represented by one of the following equations:Free energy values for the formation of plutonium trifluoride are calculated from experimentally determined equilibrium constants and the free energy of formation of uranium tetrafluoride, using the expression: The free energy of formation at 1573°K for PuF3 was found to be 93 ± 1.5 kcal/equivalent as compared to an estimated value of 94 kcal/equivalent at 1500°K based on earlier work by Brewer, Bromley, Gilles, and Lofgren.

Plutonium hexafluoride

The preparation of plutonium hexafluoride by the fluorination of plutonium trifluoride, tetrafluoride, or dioxide is described, and the effect of temperature, rate of flow of fluorine, duration of reaction, and nature of the starting material is discussed in some detail. The formula PuF 6 has been confirmed by chemical analysis. Some properties of plutonium hexafluoride are described, these include the vapour pressure, melting-point, decomposition under α-particle irradiation, and reactions with water and sulphuric acid.

Vapor Pressures of Americium Trifluoride and Plutonium Trifluoride, Heats and Free Energies of Sublimation

Vapor Pressures of Americium Trifluoride and Plutonium Trifluoride, Heats and Free Energies of Sublimation

The Melting Point and the Heat of Sublimation of Plutonium Trifluoride1

The Melting Point and the Heat of Sublimation of Plutonium Trifluoride¹

The Vapor Pressure of Plutonium Halides

Vapor pressure measurements have been made with three halides of plutonium by a modification of the Knudsen effusion method. Measurements with solid plutonium trifluoride from 1200°K to 1440°K give the vapor pressure-temperature relation log10pmm=12.468–21,120/T. Measurements with liquid plutonium trifluoride from 1440°K to 1770°K give log10pmm=11.273–19,400/T. Measurements with solid plutonium trichloride from 850°K to 1007°K give log10pmm=12.726–15,910/T; with liquid trichloride from 1007°K to 1250°K give log10pmm=9.509–12,590/T. Measurements with solid plutonium tribromide from 800°K to 929°K give log10pmm=13.386–15,280/T; with liquid tribromide from 929°K to 1100°K give log10pmm=10.321–12,360/T.