Dissolver Solution Research Articles

Extraction of palladium isotopes in irradiated actinides

A method is described for the carrier-free separation of microgram quantities of fission product palladium from dissolver solutions of irradiated actinides. The separation is effected by extracting the dimethylglyoxime (DMG) complex with chloroform. This extraction is followed by the destruction of the DMG complex with nitric and perchloric acids. Analysis of the sample is then preformed by surface ionization mass spectrometry using a modified silica gel technique. Radioactive tracers were used to follow the palladium and the most likely mass interferences, ruthenium, silver, and cadmium, through the procedure. A palladium yield of 50-60% and decontamination factors for ruthenium of >4 x 10/sup 4/, for cadmium of >10/sup 4/ and for silver of >10/sup 5/ were obtained at the microgram level. Analysis of several aliquots of a dissolver solution of irradiated /sup 239/Pu demonstrated that the achieved purification was sufficient for mass spectrometric analysis of palladium at the microgram level. 1 figure, 1 table.

Characteristics and Behavior of Emulsion at Nuclear Fuel Reprocessing

The characteristics and behavior of the emulsion formed in mixer-settlers during nuclear fuel reprocessing were studied with the dissolver solution of spent fuel burned up to 28 000 MWd/MTU and a p...

Evaluation of Inertial Filter for Dissolver Solution Clarification

Abstract The performance of some types of solvent extraction contactors is adversely affected by a high concentration of entrained solids in the process feed streams. Therefore, during reprocessing of spent reactor fuels it is desirable to separate undissolved solids and insoluble fission product residues from dissolver solutions. A series of statistically designed filtration tests were conducted to evaluate the separation efficiency of a Mott inertial filter for the removal of sub-micron particulate matter from dissolver slurries following centrifugation. Slurries used for testing consisted of 0.1 weight percent solids (needle-shaped yellow Fe2O3, rhombehedral-shaped red Fe2O3, or SiC hulls from HTGR fuel spheres) dispersed in water. Particle diameters ranged from less than 0.1 micron to 10 microns. The operating conditions varied temperature, inlet and outlet pressures, backwash intervals and particle shape. Measured separation efficiencies were greater than 90% indicating that all particles larger than...

Separation and Purification of Magnesium, Lead, and Neodymium from Dissolver Solution of Irradiated Fuel

Abstract Magnesium and lead have been identified as tracers for the input accountability of plutonium in reprocessing plants where'as fission product 148Nd is used as a burn-up monitor. Ion exchange procedures developed for the separation and purification of Mg, Pb, and Nd from the dissolver solution of irradiated fuel for their mass spectrometric analyses are presented. Distribution ratio curves of Pb, Ce, Nd, and Sm in CH3OH-HNO3-H2O medium used in the development of the separation and purification procedure are also shown.

Isotope Dilution Alpha Spectrometry for the Determination of Plutonium Concentration in Irradiated Fuel Dissolver Solution: IDAS and R-IDAS

Isotope Dilution Alpha Spectrometry (IDAS) and Reverse Isotope Dilution Alpha Spectrometry (RIDAS) have been developed for determining the concentration of plutonium in the irradiated fuel dissolver solution. The method exploits 238Pu in IDAS and 239Pu in R-IDAS as a spike and provides an alternative method in the event of nonavailability of 242Pu, which is required in Isotope Dilution Mass Spectrometry (IDMS). Depending upon the burnup of the fuel, 238Pu or 239Pu is used as a spike to change the 238Pu/(239Pu + 240Pu) alpha activity ratio in the sample by a factor of 10. This change is determined by alpha spectrometry on electro-deposited sources using a solid-state silicon surface barrier detector coupled to a multichannel analyzer. The validity of a simple method based on the geometric progression decrease for the far tail of the spectrum to correct for the tail contribution of 238Pu (5.50-MeV) peak to the low energy 239Pu + 240Pu (5.17-MeV) peak is established. Results for the plutonium concentration on different irradiated fuel dissolver solutions with burnup ranging from 1000 to 100 000 MWd/tU are presented and compared with those obtained by IDMS. The values obtained by IDAS or R-IDAS and IDMS agree to within 0.5%.

Simultaneous Plutonium and Uranium Isotopic Analysis from a Single Resin Bead - A Simplified Chemical Technique for Assaying Spent Reactor Fuels

Abstract The method uses basic anion resin to adsorb plutonium and uranium from 7–8 M HNO3 solutions containing dissolved spent reactor fuels. After equilibrating the resin with the solution, a single bead is used to determine the isotopic composition of plutonium and uranium on sample sizes as small as 10−9 to 10−10 g of each element per bead. Isotopic measurements are essentially free of isobaric interferences and fission product contamination in the mass spectrometer is eliminated. A very small aliquot of dissolver solution containing 10−6 g of U and 10−8 g of Pu is sufficient sample for chemically preparing several resin beads. A single prepared bead is loaded onto a rhenium filament and analyzed in a two-stage mass spectrometer using pulse counting for ion detection to obtain the high sensitivity required. Total quantity of the elements, in addition to isotopic abundances, can be determined by isotope dilution. Other areas where the method may be useful are: in plutonium production, isotope separations, and for trace detection of contamination on reactor parts.

Separation of Plutonium-238 from Fission Products by Solvent Extraction Using HDEHP

A procedure for separating 238Pu from a Np sample irradiated with neutrons is described. Rapid separation of Pu by HDEHP solvent extraction was attempted, and without adjusting its valency states in the dissolver solution of the sample. Both Pu(IV) and Pu(VI) were extracted along with Np from the HNO3 solutions of various concentrations. The Pu and Np extracted in the organic solution were back-extracted with oxalic acid solutions. The decontamination factors of the crude products were of the order of 102 for gross γ-activity. The Pu in the products was separated from Np by means of ion exchange resin columns. Approximately 0.5 mg of 238Pu was obtained with an efficiency exceeding 95%.

Nitrofluor法によるコロンブ石中のニオブとタンタルの抽出

The Nitrofluor process provides a thorough method of dissolving the columbite in a nonaqueous inorganic solvent, purifying the niobium and tantalum, and separating them from each other by volatile separation tecchniques.HF solutions of N2O3, N2O4 or NOF can disintegrate the columbite even at a room temperature. Subsequently, most of the niobium and tantalum precipitate as complex fluorides, leaving in the solution relatively large quantities of impurities, such as iron, manganese, titanium and tin.For further isolation of the complex fluorides of niobium and tantalum from the impurities, a volatilization of the niobium and tantalum from the dissolver solid separated from the dissolver solution is found effective. There appears to be a possibility of separating niobium and tantalum from each other by fractional sublimation utilizing the difference of volatility of these complex fluorides when the HF-N2O3 azeotrope is used for disintegrating the columbite.A general flowsheet is proposed for treating typical columbite or tantalite by the Nitrofluor process.

The x-ray spectrometric determination of uranium and plutonium in solutions of spent nuclear fuels

X-Ray spectrometry is applied to the determination of uranium and plutonium in dissolver solutions containing fission-product activities up to 1330 mCi. Uranium was measured in the concentration range 0.05–18 mgg of solution and plutonium in the range 0.6–1.2 mgg of solution with thorium as an internal standard. Plutonium was determined in the presence of a large excess of uranium and solutions with U/Pu ratios up to 300 were measured. The relative standard deviation of a single measurement was 1.0–3.0% over the concentration ranges investigated. The effect of direct γ-radiation from the sample on the detector background was demonstrated and minimized by means of lead shielding. It is suggested that samples with radioactivity levels up to 2 Ci could be measured safely and without seriously increasing the detector background. The application of this method at certain points in the nuclear fuel cycle may be important in the safeguarding of nuclear fuel.

Effect of Metallurgical Composition of Aluminum-Uranium-Silicon Fuels on Chemical Reprocessing

More economical methods of manufacturing aluminum-uranium fuel elements used extensively in high thermal-neutron flux reactors might be employed if the quantity of silicon allowed in the fuels could be substantially increased. Since silicon has created problems in reprocessing this type of fuel, various core and cladding alloys were examined for the effect of the silicon content upon dissolution, extraction, and solids production during dissolution. Dissolution rates in nitric acid were related to the metallurgical compositions of the alloys which, in turn, were related to their silicon content, but the effect was not sufficient to interfere with the reprocessing of the fuels. Emulsion stabilization, a problem during liquid-liquid extraction of uranium, is caused by the silicon that dissolves from the alloy as the intermetallic phase U(Al,Si)3. Silicon present in the alloy at a silicon:uranium atom ratio <1 is in the intermetallic fuel particle, while silicon in excess of approximately one atom per atom of uranium does not enter into the intermetallic phase and is present in the alloys as segregations of elemental silicon. This elemental silicon does not dissolve and remains in the dissolver solution as undesirable solids.

Controlled-Potential Coulometric Determination of Uranium(VI) in Uranium-Niobium Alloys.

Two procedures were developed, one in which U is determined in the presence of Nb and a second in which U is separated from Nb and then titrated. The first procedure can be used for fairly pure U-Nb solutions in which the Nb: U weight ratio is 3: 1 or less. The second procedure is used for solutions containing Nb: U weight ratios greater than 3; 1 or for solutions containing contaminants, such as Fe, that interfere in the subsequent U titration. Both methods are precise, accurate, and can be applied to the analysis of samples in which the Nb: U weight ratio is much greater than presently encountered dissolver solutions of this type. (P.C.H.)

Determination of uranium(vi) by tri- n-octylphosphine oxide extraction and coulometric titration

Uranium can be determined in the usual types of dissolver solutions by extraction of uranium (VI) into a cyclohexane solution of tri- n-octylphosphine oxide (TOPO), back-extraction into an ammonium sulfate solution, and coulometric titration at controlled potential. Optimum conditions were established for the extraction and back-extraction, and the overall perfomance of the method was evaluated. The method is accurate, precise, and widely applicable. It should be very useful in nuclear reactor technology.

Estimation of ruthenium species in solutions arising from the aqueous processing of irradiated uranium

AbstractThe estimation, by paper chromatography and solvent extraction, of the different types of nitrosylruthenium complex present in solutions of irradiated uranium is described. The ruthenium in certain ‘dissolver solutions’ has been found to occur in approximately equal proportions of (RuNO) nitrato (non‐nitro) and (RuNO) nitro complexes. The value of the ruthenium decontamination factor observed in the first extractor of the chemical separation plant at Windscale, to which these ‘dissolver solutions’ are fed, is consistent with this composition.

The Determination of Plutonium in Irradiated Uranium Fuel Solutions by Controlled-Potential Coulometry

Controlled potential coulometry has been applied to the determination of plutonium in solutions which contain plutonium within the range 0.05 to 50 g/l. The titrations are performed by direct electrode reactions at the controlled-potential electrode. No valence state adjustments or methods for detecting the end point of a titration are necessary. The electronic equipment and special cells necessary to prevent the spread of radioactivity are described in detail. The procedures are simple enough and the equipment dependable enough for routine application in control laboratories. The method has been applied to solutions containing approximately 20 g/l. of plutonium with a relative standard deviation of 0.05 per cent and with an accuracy of better than 0.1 per cent. The method has also been applied to the direct estimation of the plutonium content of solutions of irradiated natural uranium (dissolver solution) with a relative standard deviation of 0.5 per cent and an accuracy of at least 1.4 per cent, using samples containing between 100 and 200 μ g of plutonium. The titration curves follow the Nernst equation exactly for ions which undergo reversible electrode reactions. Therefore, if the concentration and half-wave potential of an interfering material are known, estimations can readily be made of the theoretical magnitude of the interference. The method is remarkably free from interferences since there are few reversibly titratable ions which have standard electrode potentials near that of the plutonium IV plutonium III couple in an acid medium. When an ion with a standard redox potential near that of the plutonium couple is known to be present, the titration can usually be carried out by the selection of a complexing medium which increases the difference between the electrode potentials of the interfering substances and that of plutonium. Several such media have been investigated and are compared. The high acid or salt content of the media used provides buffering action and prevents interference from variations in the acid content of the sample. Details concerning the procedure can be obtained from document number HW-58491. A year of experience with the method has been obtained since the publication of this document and has led to a definite preference for the titration cell using the platinum working electrode. Electrode efficiencies and reaction reproducibility have not proven to be a problem as anticipated in HW-58491. The ease of cell construction and the simplicity with which remote titrations can be performed using this cell have proved to be of great advantage.

Analytical program for processing stainless steel-uranium dioxide reactor fuel elements

Analytical programs have been described for uranium recovery processes for aluminum-uranium and zirconium-uranium reactor fuel elements. It is the purpose of this paper to describe the analytical program for the uranium recovery process involving stainless-steel-uranium-reactor fuel elements.The recovery process consists of a two-step dissolution and the regular liquid-liquid solvent extraction using TBP in kerosene. Sulfuric acid is used to dissolve the stainless steel. As a result of the sulfuric acid dissolution, the uranium is converted to a slurry. The uranium slurry is then dissolved in nitric acid. After the nitric acid dissolution, the feed material is compatible with the extraction columns used at the Idaho Chemical Processing Plant for other type fuel elements previously described.In order to analyze samples of dissolver solutions for constituents necessary for plant operation, three new analytical methods were developed. These methods are a rapid method for determining uranium, one for nitrate and one for sulfate. Methods are described for uranium, acidity, specific gravity, nitrate and sulphate in the dissolver solution, and for small amounts of uranium in the extraction column raffinates. Samples of dissolver solution and first cycle aqueous raffinate contain large amounts of fission products. These samples require shielded facilities for analytical work. The remote analytical facility described by Dykes and his co-workers was used for this work.Precision and accuracy data are given for the analytical methods described.

Recovering Uranium from Unirradiated Fuel Element Scrap

The following infomation is presented on tbe recovery of unirradiated fuel element scrap: solution techniques for U-Zr, U-staialess steel, and U-Al alloy scrap and the recovery of enriched unranium by dibutyl carbitol extraction, with 99.9% over-all efficiency. A summary flowsheet of the over-all enriched uranium scrap dissolving and purification process is shown. Schematics are included of the caustic dissolution process used on U-Al alloy scrap and the dibutyl carbitol process used to remove and purify uranium from various uranium alloy dissolver solutions. (J.H.M.)