Pu Nitrate Research Articles

The rapid photochemical reduction of U(VI) at high uranium concentrations relevant to spent nuclear fuel recycle processes

An Advanced PUREX process for the recycling of spent nuclear fuel is currently under active development in the UK. Its key aims are to avoid pure separated plutonium at all stages of the process to enhance the level of proliferation resistance, and to achieve a single cycle flowsheet that has a smaller plant footprint with consequent decreases in the capital cost and secondary wastes generated. Addressing these aims, a significant feature of the process is the co-treatment of U and Pu and thus the in situ co-conversion of mixed actinide metal nitrate solutions into oxide powders, suitable for the fabrication of new mixed metal oxide (MOx) fuel. The baseline industrial process for plutonium recovery is by oxalate precipitation; however, in order to quantitatively recover both U and Pu the uranium must be in the U(IV) oxidation state due to the high solubility of U(VI) oxalate. The first stage of this co-conversion, the development of which is reported on here, is the rapid, clean photochemical co-reduction of a mixed U(VI)/Pu(IV) nitrate stream to U(IV)/Pu(III).Here we describe a study of the reduction of U(VI) in preparation for mixed U(VI)/Pu(IV) reduction trials. Exploiting the photochemistry of U, we demonstrate the convenient and efficient photo-excitation and chemical reduction of U(VI) upon exposure to 407 nm wavelength light in the presence of alcohol-based reductants. Using a purpose built laboratory-scale photochemical reactor, U(VI) solutions of up to process-relevant concentrations of 630 mmol/dm3 (150 g/l) U have been successfully converted to a U(IV) product, achieving a conversion efficiency of ∼98% within 1.5–15 min when using propan-2-ol as a sacrificial reductant. Modelling of the dependence of the rate of U(IV) generation on initial U(VI) concentration reveals the importance of light penetration depth and effective solution mixing in determining the efficiency of the photochemical process at high U-loadings. It also reveals that the photoreduction of U(VI) to U(IV) occurs by two sequential 1-electron reductions: (i) the photochemically driven reduction of UO22+ to UO2+ by propan-2-ol, which itself is oxidised to form an α-hydroxyalkyl radical, immediately followed by (ii) a second chemical reduction of UO22+ to UO2+ and/or UO2+ to U4+ by the so-formed radical. With the addition of a nitrous acid scavenger to prevent re-oxidation of the photochemically generated U(IV), a stable product is maintained indefinitely, and the solution is suitable for subsequent oxalate co-precipitation as part of a MOx fuel fabrication process.

Localized Instantaneous Dose Rates from Inhaled Particles of 239Pu.

In this work, the authors present instantaneous local dose rates from particles of plutonium-239 oxide (239PuO) embedded in various regions of the respiratory tract. For comparison, a small number of simulations were performed in a representative region of the respiratory tract with other chemical compounds including pure metallic 239Pu, 239PuO2, 239PuO3, 239Pu2O3, and 239Pu(NO3)4. A small number of simulations were also performed with 238PuO, weapons grade Pu, and Pu from a typical radioisotope thermoelectric generator (RTG) source for the same reason. The self-shielding effect is minor for very small particles but gradually becomes more significant as the particle size increases. For particles that are 0.1 μm and larger (excluding Pu nitrate), the calculated dose rate within several microns of the particle may be sufficient to damage lung cells, but the implications of damage to such a small volume of tissue are unclear. However, it is reasonable to assume that clinical effects will be observed if a large enough volume of tissue is damaged, as might happen when large numbers of particles are inhaled. The instantaneous dose rate around a particle may be predictive of deterministic effects, scar tissue formation, and biokinetics.

The effect of feeding location on the continuous precipitation of Pu(IV) oxalate in a vortex precipitator

Abstract The effects of feeding location on crystals size and distribution, precipitation percentage of Pu(IV) oxalate were investigated in a vortex continuous precipitator when the apparent average residence time was 25 min at 50 °C. The results showed that when oxalic acid was into the free vortex zone and Pu(IV) nitrate solution was into the forced vortex one, the VMD of the particles near the outlet of the precipitator increased from 36.8 ± 1.2 µm to 45.8 ± 2.1 µm, the precipitation percentage of Pu(IV) oxalate from 97.90% ± 1.0–99.79% ± 0.05%, compared with another feeding location, both oxalic acid and Pu(IV) nitrate solution into the forced vortex zone. The two feeding methods could both prevent Pu(IV) oxalate nuclei contacting the inner wall of the precipitator during nucleation. The results showed agglomeration happened during crystals growth even under stirring. Both local supersaturations and agglomeration decided particles size distribution and the average size of Pu(IV) oxalate.

Influence of plutonyl ion on electrochemical characterization of zirconium in plutonium nitrate solutions

Abstract With the aim of evaluating the corrosion behavior of a Pu evaporator made from Zr in the Rokkasho Reprocessing Plant (RRP), the electrochemical characterization of Zr was investigated in Pu nitrate solutions. Before and after experiments, spectra of Pu nitrate solutions were recorded, and Pu4+ was oxidized to PuO2 2+ by heating these solutions. The ratio of PuO2 2+ in 3 mol dm−3 HNO3 was higher than that in 7 mol dm−3 HNO3 in our experiments. The maximum open circuit potential of Zr in the Pu nitrate solution was approximately 1 V in the Pu nitrate solution containing 7 mol dm−3 HNO3. However, there were no significant changes at high PuO2 2+ concentrations, and this value remained below the passivity breakdown potential under our experimental conditions. In this study, Zr showed high corrosion resistance, even though many PuO2 2+ ions existed in the Pu nitrate solution containing 7 mol dm−3 HNO3.

Exploring the Coordination of Plutonium and Mixed Plutonyl-Uranyl Complexes of Imidodiphosphinates.

The coordination chemistry of plutonium(IV) and plutonium(VI) with the complexing agents tetraphenyl and tetra-isopropyl imidodiphosphinate (TPIP- and TIPIP-) is reported. Treatment of sodium tetraphenylimidodiphosphinate (NaTPIP) and its related counterpart with peripheral isopropyl groups (NaTIPIP) with [NBu4]2[PuIV(NO3)6] yields the respective PuIV complexes [Pu(TPIP)3(NO3)] and [Pu(TIPIP)2(NO3)2] + [PuIV(TIPIP)3(NO3)]. Similarly, the reactions of NaTPIP and NaTIPIP with a Pu(VI) nitrate solution lead to the formation of [PuO2(HTIPIP)2(H2O)][NO3]2, which incorporates a protonated bidentate TIPIP- ligand, and [PuO2(TPIP)(HTPIP)(NO3)], where the protonated HTPIP ligand is bound in a monodentate fashion. Finally, a mixed U(VI)/Pu(VI) compound, [(UO2/PuO2)(TPIP)(HTPIP)(NO3)], is reported. All these actinyl complexes remain in the +VI oxidation state in solution over several weeks. The resultant complexes have been characterized using a combination of X-ray structural studies, NMR, optical, vibrational spectroscopies, and electrospray ionization mass spectrometry. The influence of the R-group (R = phenyl or iPr) on the nature of the complex is discussed with the help of DFT studies.

Evaluation of high resolution X-ray monochromatic wavelength dispersive X-ray fluorescence instrument for determining Pu and U in nuclear reprocessing streams

An alpha-prototype benchtop instrument was developed and tested for measuring Pu and U in spent fuel dissolver tanks in a nuclear fuel reprocessing plant. The high resolution X-ray (hiRX) instrument was designed with the aim of providing a new capability to measure samples with high radiation background with selectivity, accuracy, and precision using ultra-low volumes (7 μL) of liquid. The instrument was tested with specimens from the operational reprocessing facility at Savannah River National Laboratory. The results showed that bias was <6% for Pu nitrate process solution and < 18% for Pu in U/HNO3 matrix. Instrumental precision was <1% RSD while between microcell repeatability was <6% RSD. The limit of detection for Pu in HNO3 matrix was 3 × 10−5 g/L (100 s, 35 W). This study demonstrates that hiRX technology has the potential to provide rapid and simultaneous quantification of Pu and U without sample preparation, making it suitable for process monitoring and materials control and accountability applications.

Actinide handling after wound entry with local or systemic decorporation therapy in the rat

Background: Treatment of actinide-contaminated wounds may be problematic because of contaminant physicochemical properties, dissemination and anatomical localization. This study investigates different chelation/resection protocols after contamination of rats with americium (Am) or plutonium (Pu) nitrate or mixed oxide (MOX; uranium (U), Pu oxide).Methods: Anesthetized rats were contaminated with Am or Pu nitrate (moderately soluble) or MOX (insoluble) following wounding of hind leg muscle. DTPA (diethylene triamine pentaacetic acid) treatment (30 μmol/kg) was immediate or delayed, systemic or local and combined or not with wound resection. Actinide urinary and tissue levels were measured.Results: Comparison of Pu nitrate and MOX dissemination at the wound site indicated a more heterogeneous localization of MOX particles. In all cases DTPA treatment reduced target tissue (bone, liver) activity levels even if DTPA treatment was started 7 days after contamination. Surgery alone increased urinary excretion suggesting release from the wound site but no subsequent increases in organ retention (bone, liver) were observed. The combination of surgery and DTPA increased Pu excretion and reduced tissue levels markedly.Conclusion: This rodent model of actinide wound contamination has been used to test different treatments. It provides evidence of activity release as a result of surgery that seems not to lead to increased organ retention.

Effect of intrinsic 239Pu α-radiation on its extraction with a 40% solution of Tri-n-butyl phosphate in formal n2

In extraction regeneration of spent nuclear fuel, namely, in refining of solutions containing 0.2 M Pu and more, the radiolytic effects caused by intrinsic Pu α-radiation become noticeable. To examine this effect, Pu extraction from an aqueous solution containing 0.21 M Pu(IV) nitrate and 3 M HNO3 with a 40% solution of TBP in bis(octafluoroamyl) formal, CH2[OCH2(CF2)3CF2H]2, was performed without intercycle treatments. As a result, after 2-month contact with aqueous Pu solution, the organic phase got an irradiation dose of 470 kGy. This led to a decrease in the Pu distribution ratio by a factor of 4.1 and in the phase separation rate by a factor of 1.3 and to an increase in the viscosity by 12%; the density varied within 1–3%.

Internal Contamination by Actinides After Wounding

Internal contamination by actinides following wounding may occur in nuclear fuel industry workers or subsequent to terrorist activities, causing dissemination of radioactive elements. Contamination by alpha particle emitting actinides can result in pathological effects, either local or distant from the site of entry. The objective of the present study was to develop a robust experimental approach in the rat for short- and long- term actinide contamination following wounding by incision of the skin and muscles of the hind limb. Anesthetized rats were contaminated with Mixed OXide (MOX, uranium, plutonium oxides containing 7.1% plutonium) or plutonium nitrate (Pu nitrate) following wounding by deep incision of the hind leg. Actinide excretion and tissue levels were measured as well as histological changes from 2 h to 3 mo. Humid swabs were used for rapid evaluation of contamination levels and proved to be an initial guide for contamination levels. Although the activity transferred from wound to blood is higher after contamination with a moderately soluble form of plutonium (nitrate), at 7 d most of the MOX (98%) or Pu nitrate (87%) was retained at the wound site. Rapid actinide retention in liver and bone was observed within 24 h, which increased up to 3 mo. After MOX contamination, a more rapid initial urinary excretion of americium was observed compared with plutonium. At 3 mo, around 95% of activity remained at the wound site, and excretion of Pu and Am was extremely low. This experimental approach could be applied to other situations involving contamination following wounding including rupture of the dermal, vascular, and muscle barriers.

Direct spectrophotometric analysis of low level Pu(III) in Pu(IV) nitrate of Pu precipitation process feed solution

To achieve end user's specified PuO2, con- trolling and monitoring of Pu in its fourth valency state is essential prior to the conversion of Pu-nitrate to its oxide through oxalate precipitation process. Conventional radio- metric procedure for the analysis of Pu oxidation state in Pu-nitrate solution containing trace level of Pu(III) has limitation due to oxidation of Pu(III) during the sample preparation with respect to acidity. A simple direct spec- trophotometry using an optic fiber spectrophotometer was attempted for the estimation of trace level of Pu(III), after separating the bulk amount of Pu(IV) by maintaining the sample acidity. By using a synergistic mixture of 30 % TBP and 1 M theonyl trifluoro acetone in benzene, the Pu(IV) could be removed to a level which doesn't interfere in the Pu(III) absorption.

Electrochemical and spectroscopic studies of Pu(IV) and Pu(III) in nitric acid solutions

Electrochemical and absorption spectroscopic properties of Pu(IV) and Pu(III) in nitric acid have been investigated by using cyclic voltammetry (CV) and UV–Visible spectroscopy. CV using a glassy carbon electrode suggested that the electrochemical reaction of Pu(IV) nitrate complexes were found to be a quasi-reversible reduction to Pu(III) species. The formal redox potentials (E 0) for Pu(IV)/Pu(III) couples were +0.721, +0.712, +0.706, +0.705, +0.704, 0.694, and +0.696 V (vs. Ag/AgCl) when nitric acid concentrations are 1–7 M nitric acid solutions, respectively. These results indicate that the reduction product of Pu(IV) is only Pu(III). Further details for reaction mechanism of Pu(IV) were discussed on the basis of digital simulation of the experimental cyclic voltammograms. The absorption spectroscopic properties of Pu(III) and Pu(IV) in nitric acid solutions were investigated with UV–Visible spectrophotometry. As a result, it was founds that the intensities of the characteristic absorption peaks of Pu(III) and Pu(IV) tend to decrease with increasing nitric acid concentration for 1–8 M, and the peaks positions shifted longer or shorter wavelengths depending on the complex-forming abilities of Pu(III) and Pu(IV) with an increase in the nitric acid concentration.

Role of Anions and Reaction Conditions in the Preparation of Uranium(VI), Neptunium(VI), and Plutonium(VI) Borates

U(VI), Np(VI), and Pu(VI) borates with the formula AnO(2)[B(8)O(11)(OH)(4)] (An = U, Np, Pu) have been prepared via the reactions of U(VI) nitrate, Np(VI) perchlorate, or Pu(IV) or Pu(VI) nitrate with molten boric acid. These compounds are all isotypic and consist of a linear actinyl(VI) cation, AnO(2)(2+), surrounded by BO(3) triangles and BO(4) tetrahedra to create an AnO(8) hexagonal bipyramidal environment. The actinyl bond lengths are consistent with actinide contraction across this series. The borate anions bridge between actinyl units to create sheets. Additional BO(3) triangles and BO(4) tetrahedra extend from the polyborate layers and connect these sheets together to form a three-dimensional chiral framework structure. UV-vis-NIR absorption and fluorescence spectroscopy confirms the hexavalent oxidation state in all three compounds. Bond-valence parameters are developed for Np(VI).

Preparation and characterization of dicesium tetravalent plutonium hexanitrate

Abstract In experiments on U crystallization in a dissolver solution containing Cs, there is concern that Cs and Pu(IV) nitrate complex are deposited on the UNH crystal in the dissolver solution at the time of the U crystallization. The characteristics of generation of Cs and Pu(IV) nitrate complex with dissolver solution of MOX fuel were examined. This complex was obtained as a precipitate by mixing dissolver solution of MOX fuel and CsNO3 solution, and was identified as dicesium tetravalent plutonium hexanitrate, Cs2Pu(NO3)6 by concentration analysis and XRD. The precipitate has a tendency to be generated at high HNO3 concentrations. Thermal analysis shows that the precipitate is stable below 245 °C, and a weight loss of 10.29 ± 0.23% is observed between 245 and 297 °C. This result indicates the decomposition of Cs2Pu(NO3)6 to Cs2PuO2(NO3)4. With these properties, the UNH crystal should melt at the condition between 60 and 100 °C and be separable from the Cs complex by filtration. This suggests a new method of crystal purification allowing higher decontamination of UNH crystal in the U crystallization process.

Solubility of plutonium nitrate in uranyl nitrate hexahydrate melt

At a weight fraction below 20%, Pu(IV) nitrate is isomorphically dissolved and homogeneously distributed in the uranyl nitrate hexahydrate phase, the solubility being independent of the HNO3 concentration in the mixed U-Pu melt up to the acid concentration of 7 M. In dissolving solid Pu(IV) nitrate in 1–6 M HNO3, Pu(IV) disproportionates to form Pu(III) and Pu(VI). With increasing HNO3 concentration to above 10 M, Pu(IV) passes into solution as Pu(NO3)62−, and no disproportionation is observed.

Early Inflammatory Changes in Lung Following PuO2 or Pu Nitrate Contamination in the Rat

Increases in cancer incidence and mortality among workers overexposed to alpha-emitting radionuclides, such as plutonium (Pu), have been described with lung tumors of epithelial origin being the most common. However, the question of mechanisms leading to tumor formation following inhalation of radionuclides remains. The main factor controlling tumor incidence seems to be associated with radiation dose distribution, the more homogenous the distribution, the higher the incidence of cancer. In addition, some evidence exists regarding the role of the inflammatory response as a cofactor of tumorigenesis. The present study aims to determine the early inflammatory changes following pulmonary Pu contamination with the insoluble compound, PuO2 or the moderately soluble compound, Pu nitrate. Adult male Sprague-Dawley rats were exposed to PuO2 aerosols using a nose-only inhalation procedure (Initial lung deposit 4.7-43.4 kBq), or received intratracheal administration of Pu nitrate (25 kBq). Fourteen days post contamination, rats were euthanized and bronchoalveolar lavages (BAL) carried out. Activity was measured by liquid scintillation in lungs, femurs and liver. Distribution of activity within lung compartments was also studied. Activation of alveolar macrophages was evaluated by measurement of inflammatory mediators in supernatants collected 24h after plating and intracellular acid phosphatase activity determination, as well as by CD68 immunolabelling on lungs. The higher solubility of Pu nitrate as compared to PuO2 is illustrated by higher activity deposits in skeleton and liver, and lower retention in lungs. However, at this time point, lung distribution does not vary between the 2 compounds, the majority of activity being retained in the cellular fraction of BAL, mainly macrophages. Activation of macrophages is observed in the two groups of contaminated animals, with an enhanced production of TNF-alpha, MCP-1, CINC-1 and MIP-2, and an increased acid phosphatase activity, as compared to sham-contaminated rats. The level of activation was found to be dependant on the initial lung deposit following PuO2 contamination. Our results provide evidence for an early inflammatory response following Pu lung contamination and the role of macrophages whatever the solubility of the Pu compound.

Crystal structure of a complex of Pu(VI) nitrate with triphenylphosphine oxide, [PuO2(NO3)2(OP(C6H5)3)2

Single crystals of [PuO2(NO3)2(TPPO)2] (TPPO = OPPh3) isostructural to the related compounds of uranyl and neptunyl were isolated, and the structure of this complex was determined. Contrary to the complexes [AnO2(TPPO)4](ClO4)2 studied previously, the interatomic distances and volumes of coordination polyhedra of An in these compounds somewhat decrease in the series U-Np-Pu. This difference was attributed to a change in the number of TPPO ligands in the compounds and weakening of their interaction with oxygen atoms of the AnO 2 2+ groups in passing from [AnO2(TPPO)2](ClO4)2 to [AnO2(No3)2(TPPO)2].

In vitro dissolution study of plutonium in aerosol particles from the Mayak PA: a tool for individualised dose estimates

Chronic inhalation of Pu particles during Mayak processing is a potential concern for workers. Of the many particle properties that affect individualised dose estimates, particle solubility in lung fluids can be most important. This study compares in vitro dissolution rates of several plutonium industrial compounds present at different stages of the Mayak processing cycle using three different solvents. The results are then used to develop values of absorption parameters for individual dose assessments. In this study, the dissolution rates of nitrate, oxide and mixed plutonium aerosols were determined using a serum ultrafiltrate stimulant (SUF), phagolysosomal simulant fluid and Ringer's solution, all using a static system. According to the results obtained with SUF, Pu nitrate is absorbed into the blood to a larger extent than predicted using model parameters currently applied for Mayak workers. Absorption into the blood of 21.5 vs. 3% of deposited nuclide as current model predicts results in underestimation of systemic burden and overestimation of the lung dose. These data are being used to provide improved retrospective dose assessments for inhaled plutonium aerosols.

Decorporation of plutonium by pulmonary administration of Ca-DTPA dry powder: a study in rat after lung contamination with different plutonium forms

This study evaluates the decorporation efficacy of a pulmonary administration of a new Ca-DTPA (diethylenetriaminepentaacetic acid) dry powder (18 micromol kg(-1) of body mass) after pulmonary contamination of rats with different Pu compounds. After inhalation of PuO2, a delayed intratracheal administration of DTPA cannot reduce significantly the retention of Pu in the lungs but limits its transfer in liver and skeleton. After pulmonary contamination by Pu nitrate, early insufflation of the DTPA powder appears twice as more efficient than an i.v injection of DTPA (30 micromol kg(-1)) to reduce Pu retention in the lungs and is as effective as i.v. injection to limit the extrapulmonary deposit. In contrast, a delayed administration of DTPA cannot reduce the lung or extrapulmonary retention. In conclusion, the improvement of aerodynamic properties of DTPA powder leads to an increase of DTPA amount deposited in the lungs and enhances the body decorporation.

Small‐Angle Neutron Scattering Study of Plutonium Third Phase Formation in 30% TBP/HNO3/Alkane Diluent Systems

Third phase formation in the extraction of Pu(IV) nitrate by 30% tri‐n‐butyl phosphate (TBP) dissolved in n‐dodecane or in the highly branched diluent hydrogenated polypropylene tetramer (HPT), which may also be known as 4,4 dipropyl heptane or tétrapropylène hydrogéné, was investigated through small‐angle neutron scattering (SANS) measurements. The SANS data were interpreted using the Baxter model for hard‐spheres with surface adhesion. According to this model, the increase in scattering intensity observed when increasing amounts of Pu(NO3)4 are extracted into the organic phase, is due to interactions between small reverse micelles containing three to five TBP molecules. In n‐dodecane, the micelles interact through attractive forces between their polar cores with a potential energy of up to −2.6 kBT. This strong intermicellar attraction leads to organic phase splitting with the separation of most of the solutes of the original organic phase into a distinct phase containing interspersed layers of n‐dodecane. When HPT is the diluent, the intermicellar attraction energy calculated from the SANS data is much lower, and no third phase formation is observed under comparable chemical conditions. However, when a significant amount of the initial aqueous plutonium is in the form of plutonyl ions, PuO2 2+, the critical energy potential is reached even in HPT. A potential explanation of the effect of Pu(VI) involves the formation of a plutonyl trinitrato complex.

Inert matrix kernels for actinide incineration in high temperature reactors

Inert matrix fuels (IMF) have been the subject of intense studies for the incineration of transuranic isotopes in both fast and light water thermal reactors, as well as dedicated devices such as the accelerator driven systems. The high temperature reactor (HTR) can also be considered for these applications. The experimental investigations presented here have demonstrated the feasibility of an innovative fabrication process for the production of novel Pu based kernels. Porous yttria stabilised zirconia (YSZ) kernels are produced in an inactive facility, and entered in the gloveboxes, where they are infiltrated with a Pu nitrate solution. Following drying and calcination, they are sintered to give the final product kernels. In comparison to other kernel production methods, this infiltration process yields no undesirable radioactive liquid waste streams that have to be treated or recycled in gloveboxes. The sole effluent is gaseous NO x . This can be passed over corrosion resistant filters out of the gloveboxes and thereafter treated by conventional means. The infiltration process can be extended readily to minor actinide kernels. The dilution of Pu or minor actinides in YSZ permits the incorporation of burnable poisons to maintain the reactor safety characteristics, and can also lead to improved fuel performance and higher discharge burnups.