Alpha spectroscopic analysis of actinides (Th,U and Pu) after separation from aqueous solutions by cation-exchange and liquid extraction

The radioactivity concentration of ²³⁶Pu, ²³²U and ²²⁸Th in aqueous samples has been determined by means of alpha spectroscopy after chemical separation and pre-concentration of the radionuclides by cation exchange and liquid–liquid extraction using the Chelex-100 resin and 30% TBP/dodecan, respectively. Method calibration using a ²³⁶Pu standard solution containing the daughter radionuclides results in a detector efficiency of 18% and in a chemical recovery for cation-exchange which is (30 ± 7)%, (90 ± 5)% and (20 ± 5)% for plutonium, uranium and thorium, respectively. The chemical recovery for liquid–liquid extraction is found to be (60 ± 7)%, (50 ± 5)% and (70 ± 5)%, for plutonium, uranium and thorium, respectively. The differences in the efficiencies can be ascribed to the oxidation states, the different actinides present in solution. Taking into account that the electrodeposition of the radionuclides under study is quantitative, the total method efficiency is calculated to be (18 ± 15)%, (46 ± 7)% and (15 ± 5)%, for plutonium, uranium and thorium, respectively, at the mBq concentration range. The detection limit of the alpha spectrometric system has been found to be 0.2 mBq/L, suggesting that the method could be successfully applied for the radiometric analysis of the studied radionuclides and particularly uranium in aqueous samples.     

Determination of ultratrace amounts of plutonium in high saline groundwater by inductively coupled plasma mass spectrometry

For determination of ultratrace amounts of plutonium in high saline groundwater, large-volume sampling and preconcentration are necessary. However, traditional co-precipitation methods, such as Fe(OH)₃, Ca(OH)₂–Mg(OH)₂ and hydroxide-carbonate co-precipitation, are unable to meet the requirements of preconcentration of the ultratrace plutonium in high saline groundwater. In this paper, the ultratrace plutonium in high saline groundwater was concentrated by sequential co-precipitation with MnO₂ and Fe(OH)₃, and purified by two-stage anion-exchange separation on Dowex1 × 4 resin column. Quadrupole ICP-MS was employed for the determination of ²³⁹Pu with ²⁴²Pu spiked. After co-precipitation and purification, the major matrix elements were significantly decreased to μg mL^?1 level and decontamination factor of uranium is better than 10⁶. The detection limit for ²³⁹Pu in 50 L high saline water is 2.1 × 10^?16 g L^?1. Three high saline fountain samples (total dissolved solids more than 20 g L^?1) from Dunhuang region of China were analyzed using this method. The concentrations of ²³⁹Pu in these samples were 0.48 ± 0.2 × 10^?15, 1.40 ± 0.10 × 10^?15 and 2.13 ± 0.21 × 10^?15 g L^?1 respectively. The recovery obtained for 7 pg of ²⁴²Pu spiked into real high saline-water samples was all above 70 %.     

Analysis of Th, U, Pu, and Am in radioactive metal waste using extraction chromatography

In order to analyze actinide elements in radioactive metal waste, the dissolution and chemical separation conditions were optimized. The surfaces of a type 304 stainless steel plate and of pipe waste sampled from the prototype advanced thermal reactor (Fugen) were dissolved in mixed acid solution (HNO₃:HCl:H₂O = 1:1:4). The resulting solution was evaporated to dryness and dissolved with 2 mol/dm³ of HNO₃ to prepare sample solutions. In order to analyze trivalent actinide elements in the sample solution containing a large amount of Fe(III) (>0.1 g) using TRU resin, the effect of Fe(III) concentration on the recovery of Am(III) and reduction effect of Fe(III) to Fe(II) with ascorbic acid were studied. On the basis of results of this study, chemical separation scheme was constructed and Pu and Am in the sample solutions were separated. Thorium and U in the sample solutions were separated with UTEVA resin. High recoveries for all experimented elements were obtained from the analysis of spiked sample solutions, the effectiveness of the method was confirmed.     

Selective separation of radium and uranium from aqueous solutions by Chelex-100

The affinity of Chelex-100 for radium has been investigated as a function of pH and salinity compared to the Chelex-100 affinity for uranium to assess possible application of the resin for the selective separation of the two naturally occurring radionuclides from aqueous solutions. According to the experimental data the maximum chemical recovery of Chelex-100 is observed for uranium at pH 5 and for radium at pH 3 indicating a pH controlled selectivity of the resin for the two radionuclides. Moreover, the effect of salinity on the chemical recovery of radium is significant, resulting in a dramatic decrease of the former with increasing salinity. On the other hand, there is almost no effect of the salinity on the chemical recovery of uranium, indicating the higher affinity of Chelex-100 for uranium, which could be attributed to the formation of inner-sphere complexes of U(VI) with the iminoacetic moieties of the resin. The method has been successfully applied for the uranium separation from a radionuclide mixture.     

Determination of 210Po and uranium in high salinity water samples

A method for the determination of uranium and ²¹⁰Po in high salinity water samples has been elaborated. Both radionuclides are preconcentrated from 0.5 dm³ saline media by co-precipitation with hydrated manganese dioxide, followed by dissolution of the precipitate in 200 mL of 1 M HCl. Uranium isotopes ²³⁵U and ²³⁸U can be directly determined by ICP MS method with a detection limit of 0.01 ppb for ²³⁸U. Prior to a selective determination of ²¹⁰Po, the majority of other naturally occurring α-emitting radionuclides (uranium, thorium and protactinium) can be stripped from this solution by their extraction with a 50% solution of HDEHP in toluene. Finally, ²¹⁰Po is simply separated by direct transfer to an extractive scintillator containing 5% of trioctylphosphine oxide in Ultima Gold F cocktail and determined by an α/β separation liquid scintillation technique with detection limit below 0.1 mBq/dm³.     

New approaches to processing and fabrication of oxide nuclear fuels

It was shown that, in contrast to the Purex process using aggressive and environmentally hazardous 8M HNO₃ solutions for dissolving spent oxide nuclear fuel (SNF), this fuel can be easily dissolved in aqueous subacid ([H⁺] ∼0.1 M) solutions of Fe(III) nitrate (chloride) with partial separation of uranium and plutonium from fission products (FP). The low acidity of the solutions obtained (pH ∼1) allows direct application of modern technologies of finishing processing of nuclear fuel by fluoride, carbonate, oxalate, or peroxide precipitation of uranium and plutonium. It was established that U(VI) is isolated from nearly neutral nitric acid solutions as a poorly soluble uranyl hydroxylaminate complex after adding hydroxylamine. It was shown that on thermal decomposition at 200–300°C under ambient atmosphere this compound converts into uranium dioxide. A similar approach was applied to obtain mixed oxide uranium-plutonium fuel (MOX fuel).     

Separation of neptunium, plutonium, americium and curium from uranium with di-(2-ethylhexyl)-phosphoric acid (HDEHP) for radiometric and ICP-MS analysis

The possibility of using di-(2-ethylhexyl)-phosphoric acid (HDEHP) in solvent extraction for the separation of neptunium, plutonium, americium and curium from large amounts of uranium was studied. Neptunium, plutonium, americium and curium (as well as uranium) were extracted from HNO₃, whereafter americium and curium were back-extracted with 5M HNO₃. Thereafter was neptunium back-extracted in 1M HNO₃ containing hydroxylamine hydronitrate. Finally, plutonium was back-extracted in 3M HCl containing Ti(III). The method separates²³⁸Pu from²⁴¹Am for α-spectroscopy. For ICP-MS analysis, the interferences from²³⁸U are eliminated: tailing from²³⁸U, for analysis of²³⁷Np, and the interference of²³⁸UH⁺ for analysis of²³⁹Pu. The method has been used for the analysis of actinides in samples from a spent nuclear fuel leaching and radionuclide transport experiment.     

Radiological measurement of phosphate rock and phosphogypsum from fertilizer industries in India

Phosphate deposits are generally characterized by high levels of natural radionuclide concentrations. Natural radionuclides from the uranium and thorium series were measured, using high-resolution gamma-spectrometry in phosphate rock and phosphogypsum samples from the phosphate fertilizer industry in India. Equilibrium was found to be disrupted during the chemical processing of phosphate rock with 83 % of the ²²⁶Ra and only 5 % of ²³⁸U fractionating to phosphogypsum. Activity concentrations of ²³⁸U and ²²⁶Ra in phosphogypsum produced from various fertilizer industries of India showed levels < 1,000 Bq kg^?1 and pose no restriction for use in building/construction material.     

Separation of some radionuclides from fission product elements by adsorption on diethylene glycol succinate

Separation of adsorbed radionuclides on diethylene glycol succinate (DGS) by diethylenetriaminepentaacetic acid (DTPA) was quantitatively performed. Adsorption between the radionuclides studied and DGS was attained at pH 5±0.5. Individual separations of⁶⁰cobalt (II),⁹⁰strontium (II),¹⁴⁴cerium (III),²³³uranium (VI) and²³⁹plutonium (IV) were done by 25 ml of 10⁻⁴M of DTPA, as eluting agent, at different pH values.     

Radon (222Rn) in underground drinking water supplies of the Southern Greater Poland Region

Activity concentration of the ²²²Rn radionuclide was determined in drinking water samples from the Sothern Greater Poland region by liquid scintillation technique. The measured values ranged from 0.42 to 10.52 Bq/dm³ with the geometric mean value of 1.92 Bq/dm³. The calculated average annual effective doses from ingestion with water and inhalation of this radionuclide escaping from water were 1.15 and 11.8 μSv, respectively. Therefore, it should be underlined that, generally, it’s not the ingestion of natural radionuclides with water but inhalation of the radon escaping from water which is a substantial part of the radiological hazard due to the presence of the natural radionuclides from the uranium and thorium series in the drinking water.     

Modeling Speciation and Solubility in Aqueous Systems Containing U(IV,VI), Np(IV,V, VI), Pu(III,IV, V,VI), Am(III), and Cm(III)

A comprehensive thermodynamic model, referred to as the Mixed-Solvent Electrolyte model, has been applied to calculate phase equilibria and chemical speciation in selected aqueous actinide systems. The solution chemistry of U(IV, VI), Np(IV, V, VI), Pu(III, IV, V, VI), Am(III), and Cm(III) has been analyzed to develop the parameters of the model. These parameters include the standard-state thermochemical properties of aqueous and solid actinide species as well as the ion interaction parameters that reflect the solution’s nonideality. The model reproduces the solubility behavior and accurately predicts the formation of competing solid phases as a function of pH (from 0 to 14 and higher), temperature (up to 573 K), partial pressure of CO₂ (up to \( p_{{{\text{CO}}_{2} }} \)  = 1 bar), and concentrations of acids (to 127 mol·kg^?1), bases (to 18 mol·kg^?1), carbonates (to 6 mol·kg^?1) and other ionic components (i.e., Na⁺, Ca²⁺, Mg²⁺, OH^?, Cl^?, \( {\text{ClO}}_{4}^{ - } \), and \( {\text{NO}}_{3}^{ - } \)). Redox effects on solubility and speciation have been incorporated into the model, as exemplified by the reductive and oxidative dissolution of Np(VI) and Pu(IV) solids, respectively. Thus, the model can be used to elucidate the phase and chemical equilibria for radionuclides in natural aquatic systems or in nuclear waste repository environments as a function of environmental conditions. Additionally, the model has been applied to systems relevant to nuclear fuel processing, in which nitric acid and nitrate salts of plutonium and uranium are present at high concentrations. The model reproduces speciation and solubility in the U(VI) + HNO₃ + H₂O and Pu(IV, VI) + HNO₃ + H₂O systems up to very high nitric acid concentrations (\( x_{{{\text{HNO}}_{3} }} \approx 0.70 \)). Furthermore, the similarities and differences in the solubility behavior of the actinides have been analyzed in terms of aqueous speciation.     

Assay of actinides in human urine by rapid method

A method using DGA resin (N,N,N′,N′-tetra-n-octyldiglycolamide on an inert support) was developed for the rapid analysis of actinides in urine samples. Samples acidified with HCl to 4 M were loaded directly (without digestion) onto a DGA column. Actinides were stripped simultaneously, α-sources were prepared by co-precipitation with NdF₃. Americium, plutonium and uranium were separated with acceptable high recoveries (40–80%). The americium, plutonium and uranium content of 100–200 ml urine samples was determined within 24 h with detection limits as low as 0.01 Bq l^?1. Based on model experiments using ¹⁴C-spiked urea, it was proven that high urea content can affect americium separation deleteriously due to irreversible fixing of americium on DGA resin.     

Extrapolation studies on adsorption of thorium and uranium at different solution compositions on soil sediments

Study on adsorption of thorium and uranium radionuclides by a soil sediment as a function of ionic composition of Ca, Mg and Na has been carried out. Experimentally determined slopes represents an average of adsorption on soil sediments having different relative affinities for thorium, uranium, calcium and magnesium. Both thorium and uranium were found to be adsorbed to ion-exchange sites together with calcium and magnesium cations as effective competitors An extrapolated equation for the distribution coefficientK _d was formed for both radionuclides thorium and uranium at the specified site where the soil sediments were sampled. The combined cation concentration of both calcium and magnesium in solution correlates linearly with the measuredK _d ^Th,U values.     

Development of cost effective method for production of 64Cu from natNi

A ^natNi foil was used for the production of ⁶⁴Cu via ⁶⁴Ni(p,n)⁶⁴Cu nuclear reaction when the necessary investment for target material (350 mg) is 50 times less using the ^natNi instead of ⁶⁴Ni. The produced 32.2 ± 1.8 MBq of “no carrier added” ⁶⁴Cu is sufficient for 10 mice trials on small animal PET. The radionuclide contamination was <13 ± 12 kBq for ⁵⁵Co and 4 ± 2 kBq for ⁵⁷Ni comparing to minimum detectable activity and only 52 ± 2 kBq of ⁶¹Cu was in ⁶⁴Cu due to the modified ion exchange separation. The concentration of Fe(III) was maintained under 1.7 ppm by precipitation and filtering of Fe(OH)₃ due to the chemical purity was required.     

Complex Formation Equilibria of Unusual Seven Coordinate Fe(III) Complexes with DNA Constituents

Seven-coordinate Fe(III) complexes [Fe(dapsox)(H₂O)₂]⁺, where [dapsox = 2,6-diacetylpyridine-bis(semioxamazide)] is an equatorial pentadentate ligand with five donor atoms (2O and 3N), were studied with regard to their acid–base properties and complex formation equilibria. Stability constants of the complexes and the pK _a values of the ligands were measured by potentiometric titration. The interaction of [Fe(dapsox)(H₂O)₂]⁺ with the DNA constituents, imidazole and methylamine·HCl were investigated at 25 °C and ionic strength 0.1 mol·dm^?3 NaNO₃. The hydrolysis constants of the [Fe(dapsox)(H₂O)₂]⁺ cation (pK _a1 = 5.94 and pK _a2 = 9.04), the induced ionization of the amide bond and the formation constants of the complexes formed in solution were calculated using the nonlinear least-squares program MINIQUAD-75. The stoichiometry and stability constants for the complexes formed are reported. The results show the formation of 1:1 and 1:2 complexes with DNA constituents supporting the hepta-coordination mode of Fe(III). The concentration distributions of the various complex species were evaluated as a function of pH. The thermodynamic parameters ΔH° and ΔS° calculated from the temperature dependence of the equilibrium constants were investigated for interaction of [Fe(dapsox)(H₂O)₂] with uridine.     

Potentiometric determination of uranium in the presence of plutonium in H2SO4 medium

The potentiometric determination of uranium is widely carried out in phosphoric acid medium to suppress the interferences of plutonium by complexation. Owing to the complexity of the recycling plutonium from the phosphate based waste involving manifold stages of separation, a method has been proposed in the present paper which does not use phosphoric acid. Uranium and plutonium are reduced to U/IV/ and Pu/III/ in 1M H₂SO₄ by Ti/III/, and NaNO₂ is chosen to selectively oxidize Pu/III/ and the excess of Ti/III/. The unreacted NaNO₂ is destroyed by sulphamic acid and excess Fe/III/ is added following dilution. The equivalent amount of Fe/II/ thus liberated is titrated against standard K₂Cr₂O₇. R.S.D. obtained for the determination of uranium /1–2 mg/ is 0.3% with plutonium being present upto 4.0 mg.     

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Extraction of uranium and transuranium elements with <Emphasis Type="Italic">tert</Emphasis>-butylthiacalix[4]arene from carbonate-alkaline solutions

Extraction efficiency of uranium and transuranium elements (Np, Pu, Am and Cm) with tert-butylthiacalix[4]arene TCA from carbonate-alkaline solutions is studied and compared with that of europium (III). Plutonium (III, IV) extraction efficiency with TCA is found to be lower comparing with that of trivalent americium and europium. Extraction efficiency of studied radionuclides decreases as following: Am ? Eu ? Pu (III), U(VI), Np (V) > Pu (IV) at pH 12. Carbonate concentration increase in aqueous phase suppresses significantly extraction of all studied radionuclides, except americium. This condition can be used for americium individual recovery from complex radioactive carbonate-alkaline solutions.     

Titrimetric determination of uranium in the presence of plutonium in sulphuric acid medium

Present work summairzes a method for the estimation of uranium in the presence of plutonium involving the reduction of uranium to U/IV/ and plutonium to Pu/III/ by Zn/Hg/ followed by the selective oxidation of Pu/III/to Pu/IV/with HNO₃ catalyzed by molybdate in the presence of large sulphate concenration [5M H₂SO₄+1.5M /NH₄/₂SO₄]. The oxidation of U/IV/ by K₂Cr₂O₇ is then carried out in the presence of excess of Fe/III/ and Al/NO₃/₃ to a sharp potentiometric end point. R.S.D. obtained for 20 determinations of uranium /3–6 mg/ was 0.3% in the presence of 0.35 mg of plutonium. Larger quantity for plutonium was found to interfere.     

The application of <Emphasis Type="Italic">N</Emphasis>,<Emphasis Type="Italic">N</Emphasis>-dimethylhydroxylamine as reductant for the separation of plutonium from uranium

Both single stage and multi-stages experiments on stripping plutonium with N,N-dimethylhydroxylamine (DMHAN) as reductant with methylhydrozine (MMH) as supporting reductant were carried out. The effect of contact time, temperature, acidity, concentration of DMHAN on back-extraction rate of plutonium was investigated in the single stage experiment. The results demonstrated that the reaction of stripping Pu(IV) in the organic phase (30% TBP–kerosene) 1BF solutions by DMHAN exhibits excellent stripping efficiency. Under the given conditions, the back-extraction rate of plutonium reaches 90% within 2 min. Higher temperature, lower acidity and the increased concentration of DMHAN benifit the stripping reaction. The concentration profile of HNO₃, uranium and plutonium were determined in a multi-stages mixer-settler after the steady state of the back-extraction, and the multi-stages results show that the plutonium can be separated effectively from uranium. The recovery of plutonium and uranium reach 99.995% or over 99.99% respectively. The separation factor of U from Pu (SF_Pu/U) is about 2 × 10⁴.