Plutonium Solutions Research Articles

Specific features of reduction of plutonium(VI) ozonation products in solutions of various nature alkalis

The decomposition processes of plutonium hydroxo compounds formed under ozonation conditions in MOH (M = Li, Na, K) solutions of various concentrations were studied by UV–Vis spectroscopy using a modified nonlinear least squares method. The influence of the nature of alkali on the kinetics and mechanisms of spontaneous reduction of alkaline solutions of hydroxo compounds of plutonium(VII) was discovered. This influence and the “anomalies” in the UV–Vis spectra for ozonized plutonium solutions are associated with the presence in the systems of iron compounds in the form of impurities in commercially available LiOH, NaOH and KOH (analytically pure, chemically pure, and ultrapure grade). Even trace amounts of impurities in alkaline solutions of plutonium compounds change the mechanisms of their reduction through the active participation of iron in redox processes. They include the oxidation of iron to ferrate(VI) ions FeO42–, followed by reduction to Fe3+, probably through the stage of formation of an intermediate with a hydroxo derivative of plutonium(VI). As a result of the analysis of large arrays of spectral data, the spectra of individual components corresponding to compounds of plutonium(VI, VII) and iron (VI) were isolated.

System optimization study of hybrid K-edge/XRF densitometer for uranium–plutonium solution measurement

In this paper, a comprehensive hybrid K-edge/XRF densitometer (HKED) device model is constructed using MCNP simulation. After the modeling process, a systematic simulation study is conducted to analyze the physical parameters and material selection of KED and XRF. The simulation results reveal that the optimal parameters for the X-ray tube are an X-ray source voltage of 160 kV and a 1 mm Fe filter. The sample should be placed in a vial with an inner diameter of 1.4 cm and an outer diameter of 2 cm. For the KED technique, the determined main parameters are a 1.9 cm Fe filter rod and an inner diameter of 0.08 cm for the collimator. For the XRF technique, the determined main parameters are a 0.01 cm Gd filter and an inner diameter of 0.3 cm for the collimator, with a detector angle of 150°. After selecting appropriate parameters, the average calibration factor Δμ of the KED technique was found to be 3.301 cm2 g−1, with a relative standard deviation (RSD) of 3.36%. Additionally, the comparison between the simulated and calculated values of uranium concentration revealed a minimum measurement error of 0.4%. The minimum detection concentration of KED for uranium solutions is approximately 1 g/L. For plutonium solutions ranging from 0.5 to 20 g/L, linear fitting of the Ka1 net peak area and plutonium concentration showed a coefficient of determination (R2) of 0.999. The detection limit of XRF for plutonium measurement was 2.33✕10−4 g/L. The linear fitting coefficients (R2) of uranium concentration versus K-edge transmission rate and plutonium concentration versus Ka1 net peak area for the hybrid technique in measuring uranium-plutonium mixed solutions are determined as 0.999 and 0.996, respectively, demonstrating the response relationship of the HKED device to uranium and plutonium under different concentrations.

Mathematical and computational models for simulating transient nuclear criticality excursions within plutonium nitrate systems

This paper presents a novel methodology for the analysis of transient nuclear criticality in aqueous solutions of plutonium nitrate. The transient methodology includes one-dimensional thermal hydraulics and a detailed description of radiolytic gas formation (including aqueous concentrations of radiolysis species and bubble population, nucleation radius and advection velocity). The positive temperature feedback effect in dilute plutonium solutions, due to a low lying resonance (at ≈ 0.3 eV) in the microscopic fission cross section, is investigated. The solutions’ nuclear properties, such as intrinsic neutron source and mean neutron generation time; thermophysical properties and heat transfer coefficients; linear energy transfer and dynamics of radiolytic gas nucleation; and subcritical power, are also investigated. The effect of step and ramp reactivity insertions on transient nuclear criticality behaviour is examined, alongside the effect of fuel solution ingress and egress from the system. For a solution containing 16.0 g L−1 plutonium with an acidity of 0.2 mol L−1, increasing the width of the prescribed step reactivity insertion from 2.25 to 2.50 s increased the magnitude of the initial power peak, and decreased the time to boil, by an order of magnitude. It was observed that the characteristics of the radiolytic gas were dependent on the magnitude of the initial power peak. For example, the transients with the highest initial power peak produced an initially large quantity of radiolytic gas, with defined oscillations in their quantity (and other properties) as the transient progressed. For transients with lower power peaks, the oscillations became less defined, with a reduced period and magnitude, due to the lower quantity of radiolytic gas produced due to the initial power peak. Many of the presented simulations of nuclear criticality transients were terminated when the solution started to boil, a phenomenon which is not included in the presented methodology. Boiling occurred when the initial reactivity input induced a power spike that raised the solution temperature enough for the positive reactivity feedback effect to sustain the transient even when the reactivity input was removed.

The Plutonium Temperature Effect Program

Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect, but up to now, no experimental program has confirmed this effect. The French Plutonium Temperature Effect Experimental Program (or PU+ in short) aims to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The PU+ experiments were conducted in the “Apparatus B” facility at the CEA VALDUC research center in France. It involved several sub-critical approach-type experiments using plutonium nitrate solutions with concentrations of 14.3, 15, and 20 g/L at temperatures ranging from 20 to 40 °C. Fourteen (five at 20 g/L, four at 15 g/L, and five at 14.3 g/L) phase I experiments (consisting of independent sub-critical approaches) were performed between 2006 and 2007. The impact of the uncertainties on solution acidity and plutonium concentration made it difficult to demonstrate the positive temperature effect, requiring an additional phase II experiment (with a unique plutonium solution) from 22 to 28 °C that was performed in July 2007. This phase II experiment has shown the existence of a positive temperature effect of ~+5.17 pcm/°C (from 22 to 28 °C for a plutonium concentration of 14.3 g/L). It has recently been possible to confirm the results of this program with MORET 5 calculations by generating thermal scattering data S(α,β) at the correct experimental temperatures. This paper finally presents a fully documented experimental program highlighting the Plutonium Temperature Effect theoretically described in the literature. Its high level of precision and its “one-step” approach to criticality allowed it to show a significant positive temperature effect for a rather small variation of temperature (+6 °C). The order of magnitude of the effect was confirmed with Monte Carlo calculations using thermal scattering data for hydrogen in the solution produced by IRSN for the purpose of the comparison.

Studying the Isotope Composition of Nitric Acid Solutions of Plutonium via Alpha and Gamma Spectroscopy

Studying the Isotope Composition of Nitric Acid Solutions of Plutonium via Alpha and Gamma Spectroscopy

Determination of uranium and plutonium in mixed oxide samples by X-ray methods and their comparison with potentiometry

Uranium and plutonium in nuclear fuels are generally determined by conventional potentiometric methods which generates large quantities of radioactive waste. In the present study, K-edge densitometry (KED) and Energy dispersive X-ray fluorescence (EDXRF) have been examined for the determination of uranium and plutonium in solutions as an alternative to potentiometric methods. In the present study, uranium plutonium mixed oxide (MOX) pellets were dissolved in nitric acid medium containing small quantities of HF. Uranium and plutonium in MOX solutions were determined by KED and EDXRF methods and the results are compared with the one obtained by conventional potentiometric methods. The relative standard deviation (RSD) of the analytical results for U and Pu using KED were found to be better than ±1% and ± 0.9% respectively. Similarly, RSD for U and Pu by EDXRF were found to be better than ±1.6% and ± 0.8%, respectively. The absolute difference between KED and EDXRF with respect to potentiometric data for uranium and plutonium were calculated and it was found to be marginal indicating that KED and EDXRF methods can be employed for the routine analysis of U and Pu in solutions. X-ray methods employed in the present study are non-consumptive, direct and offer precise determination of U and Pu and generate no waste.

Improvement of the 241Pu/239Pu isotope ratio measurement by thermal ionisation mass spectrometry - An approach to account for the 241Am in-growth after plutonium purification

Reliable determinations of plutonium isotope ratios are required in the nuclear field. Thermal ionisation mass spectrometry (TIMS) is used for this purpose. Commercially available TIMS instruments cannot distinguish between 241Am and 241Pu. A purification step before measurement to remove americium from plutonium solutions is required. Nevertheless, the amount of 241Am in-growth due to 241Pu decay (half-life 14.3 years) in the time between plutonium purification and TIMS measurements can be significant for an accurate determination of the 241Pu/239Pu isotope ratio. This paper presents a method, with a focus on nuclear safeguards, to correct for the 241Am in-growth after plutonium purification. In addition, the method provides an estimate for the americium ionisation efficiency to allow an even more accurate determination of the 241Pu/239Pu isotope ratio.

Plutonium assay by spectrophotometry and estimation of uncertainties on routine glove box samples

Plutonium(VI) spectrophotometry is a precise analytical method for determining the concentration of plutonium solutions. It is simple and cost-effective to implement in nuclear environments such as glove boxes and hot-cells. Uncertainties were assessed at 0.64% (k = 1) for solutions of 0.150 mg g−1 initially prepared in a glove box. The critical limit of the method was estimated at 0.002 mg g−1 at 99% confidence. The good performances of the technique in the present work are due to a rigorous spectra processing as well as preparation steps conducted by weight control.

A simple microanalytical method for trace elemental determination in plutonium samples using energy dispersive X-ray fluorescence

A simple microanalytical method for trace elemental determinations in plutonium (Pu) solution samples was developed using Energy Dispersive X- Ray Fluorescence (EDXRF) spectrometry. A very small sample volume (200 μL) of Pu solution was taken for analysis and the major matrix Pu was separated using the extractant, Tri n-Butyl Phosphate (TBP), in 4 M HNO3 medium. Aliquot of 5 μL of the aqueous phase containing only the trace elements were deposited on a thin transparent mylar film inside the PVC (Poly Vinyl Chloride) EDXRF sample container and dried in ambient atmosphere inside a radioactive fume hood. After drying, the container was sealed with PVC lids. Two pure Pu solutions spiked with different trace elements having concentration of 6.25 and 23.25 μg/mL, were analyzed with RSD (Relative Standard Deviation) of 6.6 and 4.2% (1 σ, n = 3), and the average deviation from the expected values calculated on the basis of sample preparation were 8.9 and 6.6%, respectively. The developed methodology requires small amount of sample, extractant, and therefore generates less radioanalytical waste. Moreover, due to the requirement of small amount of sample, sealed inside a PVC container, the measurements could be carried out in ambient atmosphere without putting the instrument inside glove box. Further, the simple instrumentation of EDXRF, made this methodology very much ideal for the routine analysis of plutonium based samples.

Development of a Technology for Dissolving a Homogeneous Mixture of Uranium and Plutonium Oxides Obtained by the Plasma-Chemical Method

A plasma technology for the synthesis of mixed uranium and plutonium oxides UO2–PuO2 by denitration of mixed solutions of these metals has been developed at the Siberian Chemical Combine. The production process of U and Pu oxides includes the preparation of mixed initial solutions, their denitration in a plasma reactor with an inductive high-frequency discharge, and isolation of powders from the vapor–gas flow in vortex precipitators or filters with metal–fabric meshes. The particles of the obtained oxide powders possess the features of nanocrystalline and nanoscale materials containing crystalline and amorphous phases in the structure. The equipment for the implementation of this technology has been developed (Zenith plasma plant), on which mixed U and Pu oxides have been produced, and over 60 kg of oxides UO2–PuO2 with the plutonium content of 25 wt % has been synthesized, from which industrial fuel rods for the BN-600 reactor have been fabricated. After the operation of the fuel rods in the BN-600 reactor and corresponding exposure, the cores of the irradiated fuel rods should be dissolved for subsequent radiochemical treatment. The results of the studies on the search for the conditions of dissolution and effective solvents for mixed U and Pu oxides obtained by the plasma technology of decomposition of mixed solutions of uranium and plutonium are presented. It has been found that no complete dissolution of the synthesized oxides occurs in solutions of HNO3 and HNO3 with the addition of the reducing agents for tetravalent plutonium—ascorbic acid and hydrazine. Complete dissolution of the mixed oxides most effectively occurs in solutions of 10 M HNO3 with the addition of 0.25–1.0 g/L HF.

Development and validation of hybrid K-edge/K-XRF densitometer for assay of nuclear fuel from reprocessing plants

A non-destructive and noninvasive technique is preferable for characterization of the spent fuel solutions in the reprocessing plants, especially samples associated with high radiation dose. A hybrid K-edge/KXRF densitometer (HKED) technique has been designed, fabricated, and commissioned indigenously to accomplish the above objective for the spent fuel solutions from compact reprocessing facility for advanced fuels in lead cells (CORAL), Kalpakkam. This paper describes the customized design, development, calibration and validation of the HKED technique using a series of uranium solutions and corroborating the results with potentiometric method. Further, the pure and mixture of thorium and uranium as well as uranium and plutonium solutions were also assayed by HKED technique.

Material balance evaluation of pyroprocessing for minor actinide transmutation nitride fuel

ABSTRACT Fuel cycle technology for the transmutation of long-lived minor actinides (MAs) using an accelerator-driven system has been developed under the double-strata fuel cycle concept. A mononitride solid solution of MAs and plutonium diluted with zirconium nitride is a prime fuel candidate for transmutation of MAs. Pyroprocessing is suitable for recycling the residual MAs in spent nitride fuel with high radiation doses and decay heat. The spent nitride fuel is anodically dissolved, and the actinides are recovered simultaneously into a liquid cadmium cathode via molten salt electrorefining, which is the core step of the process. The process should be designed to achieve the target recovery yield of MAs and the acceptable impurity level of rare earth fission products in the recovered material. In this study, we evaluated the material balance during the pyroprocessing of spent nitride fuel to gain important insight on the design process. We examined the effects of changing the stage numbers in the countercurrent extraction and the zeolite treatment on material flow and quantity of waste. The results show that multistage extraction is necessary to meet the targets and the stage number of the zeolite treatment is recommended to be more than three from the aspect of the amounts of glass-bonded sodalite wastes.

Direct Multielemental Trace Determinations in Plutonium Samples by Total Reflection X-ray Fluorescence Spectrometry Using a Very Small Sample Amount.

A simple, safe, and sensitive method for direct multielemental trace determinations in plutonium samples using total reflection X-ray fluorescence (TXRF) spectrometry has been developed. A very small volume (2 μL) of the sample solutions was deposited on TXRF supports after separation of the plutonium matrix from these solutions. Since the amount of the plutonium deposited on the supports was in the ng level only fixed on the supports and the specimen spots were not disturbed during the sample preparation, the samples could be analyzed directly without putting the instrument in a glovebox. This approach avoided a cumbersome operation of the instrument in a glovebox, which is normally utilized for Pu-based samples using other techniques. Similarly, the requirement of small amounts of the samples minimized the radiation dose to the operator as well as a cumbersome problem of management of radioactive analytical waste of plutonium samples. The samples were analyzed using the TXRF spectra of the specimens, concentration of the internal standard Se or Ga and predetermined sensitivity values. The elemental detection limits for the elements K-Sr varied from 1.06 to 0.09 ng. The elements K, Ca, Cr, Mn, Fe, Ni, Cu, Zn, Sr, Ba, Tl, and Pb were analyzed at μg/mL level. The analytical results of TXRF determinations showed average relative standard deviation (RSD) value of 4.5% (1σ, n = 3) and the TXRF determined results deviated from the expected values by 5.9% on average for samples prepared by adding multielements in plutonium solutions. Two real plutonium samples were also analyzed in similar manner. For the real plutonium sample solution the average RSD values of TXRF determinations were 10.6% (1σ, n = 3) for the elemental concentrations in the range of 0.2 to 61 μg/mL. These values are comparable with conventional trace element analytical techniques with added advantages mentioned above.

Molecular Hydrogen Yields from the α-Self-Radiolysis of Nitric Acid Solutions Containing Plutonium or Americium.

The yield of molecular hydrogen, as a function of nitric acid concentration, from the α-radiolysis of aerated nitric acid and its mixtures with sulfuric acid containing plutonium or americium has been investigated. Comparison of experimental measurements with predictions of a Monte Carlo radiation track chemistry model shows that, in addition to scavenging of the hydrated electron, its precursor, and the hydrogen atom, the quenching of excited state water is important in controlling the yield of molecular hydrogen. In addition, increases in solution acidity cause a significant change in the track reactions, which can be explained as resulting from scavenging of eaq- by Haq+ to form H•. Although plutonium has been shown to be an effective scavenger of precursors of molecular hydrogen below 0.1 mol dm-3 nitrate, previously reported effects of plutonium on G(H2)α between 1 and 10 mol dm-3 nitric acid were not reproduced. Modeling results suggest that plutonium is unlikely to effectively compete with nitrate ions in scavenging the precursors of molecular hydrogen at higher nitric acid concentrations, and this was confirmed by comparing molecular hydrogen yields from plutonium solutions with those from americium solutions. Finally, comparison between radionuclide, ion accelerator experiments, and model predictions leads to the conclusion that the high dose rate of accelerator studies does not significantly affect the measured molecular hydrogen yield. These reactions provide insight into the important processes for liquors common in the reprocessing of spent nuclear fuel and the storage of highly radioactive liquid waste prior to vitrification.

Accurate determination of minor isotope ratios in individual plutonium–uranium mixed particles by thermal ionization mass spectrometry

Abstract Isotopic analyses of plutonium and low-enriched uranium mixtures with particle sizes of 0.6–3.3 μm were performed using thermal ionization mass spectrometry with a continuous heating method to verify its effectiveness for the accurate analysis of minor isotopes without sample pretreatment. The mixed particles used in this study were prepared from a mixed solution of plutonium (SRM 947) and uranium (U010, 235U 1% enriched) reference materials. The isotope ratios for plutonium in the individual mixed particles, including 238Pu/239Pu, 241Pu/239Pu as well as 240Pu/239Pu, and 242Pu/239Pu, were in good agreement with the certified values despite the isobaric interference of 238U and 241Am. The isotope ratios for uranium in the mixed particles also agreed well with the certified values within the range of error. However, the isotope ratios for minor isotopes, such as 234U and 236U, in the particles with diameters of less than approximately 1.8 μm could not be measured because numbers of 234U and 236U atoms in analyzed particles are too low. These results indicate that thermal ionization mass spectrometry with a continuous heating method is applicable for the analysis of trace amounts of plutonium isotopes, including 238Pu and 241Pu, despite the presence of the respective isobars 238U and 241Am in the microsamples.

Charge conservation applied to experimental data for plutonium solutions

The principle of charge conservation in isolated systems is used to analyze data recorded for an academic thesis. The analysis reveals a gradual chemical reduction of the plutonium during studies of its behavior in dilute aqueous acids. The reduction effect is reproduced in several different studies. The effect was apparently unnoticed by the author.

Plutonium: disproportionation reactions and acidity changes

Disproportionation reactions can noticeably change the acidity of a plutonium solution. The estimation of the pH change depends on the the Pu concentration, the initial oxidation-state distribution, the initial pH of the solution, and the values of three equilibrium constants. The method can be used to estimate the minimum Pu concentration that produces a specified change in the solution acidity. The results can be checked.

Ultra-trace analysis of plutonium by thermal ionization mass spectrometry with a continuous heating technique without chemical separation

Thermal ionization mass spectrometry (TIMS) with a continuous heating technique is known as an effective method for measuring the isotope ratio in trace amounts of uranium. In this study, the analytical performance of thermal ionization mass spectrometry with a continuous heating technique was investigated using a standard plutonium solution (SRM 947). The influence of the heating rate of the evaporation filament on the precision and accuracy of the isotope ratios was examined using a plutonium solution sample at the fg level. Changing the heating rate of the evaporation filament on samples ranging from 0.1fg to 1000fg revealed that the influence of the heating rate on the precision and accuracy of the isotope ratios was slight around the heating rate range of 100–250mA/min. All of the isotope ratios of plutonium (SRM 947), 238Pu/239Pu, 240Pu/239Pu, 241Pu/239Pu and 242Pu/239Pu, were measured down to sample amounts of 70fg. The ratio of 240Pu/239Pu was measured down to a sample amount of 0.1fg, which corresponds to a PuO2 particle with a diameter of 0.2μm. Moreover, the signals of 239Pu could be detected with a sample amount of 0.03fg, which corresponds to the detection limit of 239Pu of 0.006fg as estimated by the 3-sigma criterion. 238Pu and 238U were clearly distinguished owing to the difference in the evaporation temperature between 238Pu and 238U. In addition, 241Pu and 241Am formed by the decay of 241Pu can be discriminated owing to the difference in the evaporation temperature. As a result, the ratios of 238Pu/239Pu and 241Pu/239Pu as well as 240Pu/239Pu and 242Pu/239Pu in plutonium samples could be measured by TIMS with a continuous heating technique and without any chemical separation processes.

Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH

The (U,Pu)O2 matrix behavior of an irradiated MIMAS-type (MIcronized MASter blend) MOX fuel, under radiolytic oxidation in aerated pure water at pH 5–5.5 was studied by combining chemical and radiochemical analyses of the alteration solution with Raman spectroscopy characterizations of the surface state. Two leaching experiments were performed on segments of irradiated fuel under different conditions: with or without an external γ irradiation field, over long periods (222 and 604days, respectively). The gamma irradiation field was intended to be representative of the irradiation conditions for a fuel assembly in an underwater interim storage situation.The data acquired enabled an alteration mechanism to be established, characterized by uranium (UO22+) release mainly controlled by solubility of studtite over the long-term. The massive precipitation of this phase was observed for the two experiments based on high uranium oversaturation indexes of the solution and the kinetics involved depended on the irradiation conditions. External gamma irradiation accelerated the precipitation kinetics and the uranium concentrations (2.9×10−7mol/l) were lower than for the non-irradiated reference experiment (1.4×10−5mol/l), as the quantity of hydrogen peroxide was higher. Under slightly acidic pH conditions, the formation of an oxidized UO2+x phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix. The Raman spectroscopy performed on the heterogeneous MOX fuel matrix surface, showed that the fluorite structure of the mainly UO2 phase surrounding the Pu-enriched aggregates had not been particularly impacted by any major structural change compared to the data obtained prior to leaching. For the plutonium, its behavior in solution involved a continuous release up to concentrations of approximately 3×10−6molL−1 with negligible colloid formation. This data appears to support a predominance of the +V oxidation state for plutonium in solution under highly oxidizing conditions.Furthermore, the Raman spectroscopy monitoring of the sample surface oxidation states did not point to any significant effect from the high Pu content of the aggregates (10–15%) and therefore did not indicate a better aggregate stability under radiolysis compared to the mainly UO2 matrix. This is because acidic pH conditions do not favor the development of oxidized layers on a fuel surface, with the exception of secondary phases.

Application of a hydroxylamine nitrate stability model to plutonium purification process equipment

A mathematical model that predicts hydroxylamine nitrate (HAN) (NH2OH·HNO3) stability is applied to aqueous solutions containing HAN, nitric acid and plutonium that are used in plutonium purification processes. The model estimates the stability of these solutions with respect to the rapid, hazardous, autocatalytic reaction of HAN with nitric acid that generates heat and gas. It also accounts for reaction kinetics, temperature changes, gas generation rates, solution volumes and flow rates, and distribution of plutonium and nitric acid between aqueous and organic phases. The model is applied to three typical process vessels used in solvent extraction purification of plutonium – a countercurrent aqueous/organic plutonium stripping column, an oxidation column used for HAN and hydrazine destruction, and a plutonium rework tank. Both normal and off-normal process scenarios are modeled. Two of the off-normal scenarios lead to the rapid autocatalytic reaction of HAN with nitric acid where heat and gas are generated and that could lead to damage of the process equipment and/or release of hazardous plutonium solution from the vessel. In these two cases, stationary aqueous solutions containing HAN, Pu(III), and nitric acid were allowed to slowly react until conditions for the autocatalytic reaction were reached.