IC-ICP-MS applied to the separation and determination of traces of plutonium and uranium in aqueous leachates and acid rinse solutions of UO2 doped with Pu

It is expected that spent nuclear fuel, today mainly UO₂, may become exposed to groundwater after extended storage in a deep geologic repository. After 1000 years, the radioactivity of the fuel will be constituted essentially by α-emissions by long-lived actinides. The α-emissions play a significant role in determining the dissolution behavior of uranium, because the radiolysis of water results in the formation of oxidizing chemical species near the fuel surface. In order to study this effect, UO₂ doped with 0.1 and 10 wt.% of a strong α-emitter (namely ²³⁸Pu) was subject to leaching at room temperature (RT) in deionized water. In order to study the mechanisms of leaching in simulated conditions, very precise and accurate techniques need to be employed. In this paper, the results obtained by inductively coupled plasma mass spectrometry coupled with ion chromatography for the determination of traces of ²³⁸Pu and uranium in aqueous leachates’ solutions are illustrated.     

Solubility of (UO2)3(PO4)2?4H2O in H+-Na+-OH?-H2PO?4-HPO2?4-PO3?4-H2O and Its Comparison to the Analogous PuO2 +2 System

The objectives of this study were to address uncertainties in the solubility product of (UO₂)₃(PO₄)₂⋅4H₂O(c) and in the phosphate complexes of U(VI), and more importantly to develop needed thermodynamic data for the Pu(VI)-phosphate system in order to ascertain the extent to which U(VI) and Pu(VI) behave in an analogous fashion. Thus studies were conducted on (UO₂)₃(PO₄)₂⋅4H₂O(c) and (PuO₂)₃(PO₄)₂⋅4H₂O(am) solubilities for long-equilibration periods (up to 870 days) in a wide range of pH values (2.5 to 10.5) at fixed phosphate concentrations of 0.001 and 0.01 M, and in a range of phosphate concentrations (0.0001–1.0 M) at fixed pH values of about 3.5. A combination of techniques (XRD, DTA/TG, XAS, and thermodynamic analyses) was used to characterize the reaction products. The U(VI)-phosphate data for the most part agree closely with thermodynamic data presented in Guillaumont et al.,⁽¹⁾ although we cannot verify the existence of several U(VI) hydrolyses and phosphate species and we find the reported value for formation constant of UO₂PO⁻₄ is in error by more than two orders of magnitude. A comprehensive thermodynamic model for (PuO₂)₃(PO₄)₂⋅4H₂O(am) solubility in the H⁺-Na⁺-OH⁻-Cl⁻-H₂PO⁻₄-HPO²⁻₄-PO³⁻₄-H₂O system, previously unavailable, is presented and the data shows that the U(VI)-phosphate system is an excellent analog for the Pu(VI)-phosphate system.     

Application of instrumental radioanalytical techniques to nuclear waste testing and characterization

This paper describes the validation of a multi-technique analytical methodology that uses inductively coupled plasma-mass spectrometry, α-spectrometry, and γ-spectrometry for the routine analysis of samples containing transuranic radionuclides. This methodology is capable of the determination of concentrations of both²³⁸Pu and²⁴¹Pu in the presence of²³⁸U and²⁴¹Am without the need for chemical separations. The relative merits of these three techniques were evaluated as they are applied in a nuclear waste material and spent nuclear fuel testing program by analytical (1) standards and (2) solutions prepared from the dissolution of glasses doped with²³⁷Np,²³⁹Pu, and²⁴¹Am. The uncertainty associated with technique was within ±4% for standards and ±10% for doped nuclear waste glasses. The methodology was then used to analyze three fully radioactive waste glasses.     

Determination of <Superscript>238</Superscript>Pu in plutonium bearing fuels by thermal ionization mass spectrometry

Determination of ²³⁸Pu in plutonium bearing fuels is required as a part of the chemical quality assurance of nuclear fuels. In addition, the determination of ²³⁸Pu is required in nuclear technology for many other applications, e.g., for developing isotope correlations and while using ²³⁸Pu as a spike (tracer) in isotope dilution α-spectrometry (IDAS). This determination usually involves the use of α-spectrometry on purified Pu sample. In view of the random errors associated with the counting statistics and the systematic errors due to (1) in-growth of ²⁴¹Am in purified Pu sample and (2) tail contribution correction methodology in α-spectrometry, the precision and accuracy obtainable by α-spectrometry are limited. Thermal ionization mass spectrometry (TIMS) is generally used for the determination of different Pu isotopes other than ²³⁸Pu. This is due to the ubiquitous isobaric interference from ²³⁸U at ²³⁸Pu in TIMS. Recently, we have carried out studies on the formation of atomic and oxide ions of U and Pu by TIMS and developed a novel approach using interfering element correction methodology to account for the isobaric interference of ²³⁸U at ²³⁸Pu in TIMS. This methodology is based on the addition of ²³⁵U (enrichment >90 atom%) to Pu sample followed by the determination of ²³⁸U/²³⁵U atom ratio using UO⁺ ion and determination of Pu isotope ratios using Pu⁺ ion, from the same filament loading. The TIMS methodology was used for the determination of ²³⁸Pu in different Pu samples in U based nuclear fuels from PHWRs with ²³⁸Pu content about 0.2 atom%. The ²³⁸Pu determination was also carried out using α-spectrometry. This paper reports the results obtained by the two methods and presents the ments and shortcomings of the two approaches.     

Sol-Gel Microsphere Pelletisation Process for Fabrication of (U,Pu)O2, (U,Pu)C and (U,Pu)N Fuel Pellets for the Prototype Fast Breeder Reactor in India

A “dust-free” sol-gel microsphere pelletisation (SGMP) process has been developed for fabrication of (U,Pu)O₂, (U,Pu)C and (U,Pu)N fuel pellets containing around 15% plutonium for the forthcoming prototype fast breeder reactor (PFBR) in India. The objective was to produce homogeneous sintered pellets of ∼85% T.D. with a predominantly open-pore structure. Hydrated gel-microspheres of UO₃+PuO₂ and UO₃+PuO₂+C have been prepared from nitrate solutions of uranium and plutonium by the “ammonia internal gelation” process, using hexamethylene tetramine (HMTA) as an ammonia generator and silicone oil at 90±1°C as gelation bath. For oxide fuel pellets, the hydrated UO₃+PuO₂ gel-microspheres were calcined at around 700°C in Ar+8% H₂ atmosphere to produce “non-porous”, “free-flowing” and coarse (around 400 micron) microspheres which could be directly pelletised at 550 MPa to green pellets. The mixed oxide pellets were subjected either to low temperature (∼1100°C) oxidative sintering (LTS) in N₂+air containing ∼1500 ppm O₂ or to high temperature (1650°C) sintering, (HTS) in Ar+8% H₂. For monocarbide and mononitride pellets, hydrated gel-microspheres of UO₃+PuO₂+C were subjected to carbothermic synthesis in vacuum (∼1 Pa) and flowing nitrogen (flow rate: 1.2 m³/h) in the temperature range of 1450–1550°C respectively. The monocarbide and mononitride microspheres thus produced were relatively hard and required higher compaction pressure (∼1200 MPa) for making reen pellets which could be sintered to 85% T.D. in Ar+8% H₂ at 1700°C. The sintered oxide, monocarbide and mononitride pellets had a “blackberry” “open” pore microstructure with fine grain size. The microspheres retained their individual identity in the sintered pellets because during sintering densification took place mainly within and not between the microspheres.     

Synthesis and characterization of nanometric powders of UO2+x, (Th,U)O2+x and (La,U)O2+x

This paper describes a new way of preparing nanometric powders of uranium oxide, to fit the needs of studies on UO₂ oxidation, through the electrochemical reduction of U(VI) into U(IV). These powders can also be doped with radionuclides if necessary. The precipitation of oxides occurs in reducing and anoxic conditions. This original method makes it possible to synthesize nanometric UO₂ powders with a calibrated size, as well as the Th- and La-doped UO₂ powders with a predefined composition. The powder characterization by the X-ray diffraction, X-ray photoelectron spectroscopy and transmission electron Microscopy shows the formation of spherical crystallites of UO_2+x, (Th,U)O_2+x and (La,U)O_2+x phases. The composition can be defined by the initial Th/(Th+U) and La/(La+U) ratios in solution and the particle size can be controlled by varying the pH.     

Determination of actinide nuclides in water samples from the primary circuit of a research reactor

Primary coolant samples from a research have been analyzed for^239,240Pu,²³⁸Pu,²³⁸U,²³⁷Np and²³⁹Np. The determination of²³⁷Np and²³⁸U was carried out with the help of isotope dilution neutron activation analysis with²³⁹Np or²³⁸Np as tracer. For determination of^239,240Pu and²³⁸Pu alpha spectroscopic isotope dilution analysis with²³⁸Pu as tracer was used.²³⁹Np was determined with the help of isotope dilution analysis using²³⁸Np as tracer. Nuclides were isolated by chemical separation on anionite resin. Before measurement, Pu isotopes were electrolytically deposited on stainless steel plates. Activity ratios referred to²³⁸U were reported. They are helpful for identification of the sources of actinide activity in reactor effluents.     

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.     

Dissolution Mechanism of UO2 in Nitric Acid Solution by Photochemical Reaction

Photodissolution tests of UO₂ sintered pellets were carried out in 3M nitric acid solution and at about 50 °C under UV irradiation. The light source was a Hg-lamp emitting a light of 254nm wavelength. In the products, chemicals such as H₂O₂ and NO₂ ^– ion were detected during photodissolution of the UO₂ sintered pellets. Based on this result, a new dissolution mechanism of UO₂ in nitric acid solution by photochemical reaction was suggested in this study.     

Charge distribution and local structure and speciation in the UO2+x and PuO2+x binary oxides for x?0.25

The local structure and chemical speciation of the mixed valence, fluorite-based oxides UO_2+x (0.00?x?0.20) and PuO_2+x/PuO_2+x−y(OH)_2y·zH₂O have been determined by U/Pu L_III XAFS spectroscopy. The U spectra indicate (1) that the O atoms are incorporated as oxo groups at short (1.75 Å) U-O distances consistent with U(VI) concomitant with a large range of U displacements that reduce the apparent number of U neighbors and (2) that the UO₂ fraction remains intact implying that these O defects interact to form clusters and give the heterogeneous structure consistent with the diffraction patterns. The PuO_2+x system, which does not show a separate phase at its x=0.25 endpoint, also displays (1) oxo groups at longer 1.9 Å distances consistent with Pu(V+δ), (2) a multisite Pu-O distribution even when x is close to zero indicative of the formation of stable species with H₂O and its hydrolysis products with O²⁻, and (3) a highly disordered, spectroscopically invisible Pu-Pu component. The structure and bonding in AnO_2+x are therefore more complicated than have previously been assumed and show both similarities but also distinct differences among the different elements.     

New reprocessing system for spent nuclear reactor fuel using fluoride volatility method

Our proposed spent nuclear fuel reprocessing technology named FLUOREX, which is a hybrid system using fluoride volatility and solvent extraction, meets the requirements of the future thermal/fast breeder reactors (coexistence) cycle. We have been done semi-engineering and engineering scale experiments on the fluorination of uranium, purification of UF₆, pyrohydrolysis of fluorination residues, and dissolution of pyrohydrolysis samples in order to examine technical and engineering feasibilities for implementing FLUOREX. We found that uranium in spent fuels can be selectively volatilized by fluorination in the flame type reactor, and the amount of uranium volatilized is adjusted from 90% to 98% by changing the amount of F₂ supplied to the reactor. The volatilized uranium is purified using UO₂F₂ adsorber for plutonium and purification methods such as condensation and chemical traps for fission products provide a decontamination factor of over 10⁷. Most of the fluorination residues that consist of non-volatile fluorides of uranium, plutonium, and fission products are converted to oxides by pyrohydrolysis at 600-800 °C. Although some fluorides of fission products such as alkaline earth metals and lanthanides are not converted completely and fluorine is discharged into the solution, oxides of U and Pu obtained by pyrohydrolysis are dissolved into nitric acid solution because of the low solubility of lanthanide fluorides. These results support our opinion that FLUOREX has great possibilities for being a part of the future spent nuclear fuel cycle system.     

[Ni(H2O)4]3[U(OH,H2O)(UO2)8O12(OH)3], crystal structure and comparison with uranium minerals with U3O8-type sheets

The new U(VI) compound, [Ni(H₂O)₄]₃[U(OH,H₂O)(UO₂)₈O₁₂(OH)₃], was synthesized by mild hydrothermal reaction of uranyl and nickel nitrates. The crystal-structure was solved in the P-1 space group, a=8.627(2), b=10.566(2), c=12.091(4) Å and α=110.59(1), β=102.96(2), γ=105.50(1)°, R=0.0539 and wR=0.0464 from 3441 unique observed reflections and 151 parameters. The structure of the title compound is built from sheets of uranium polyhedra closely related to that in β-U₃O₈. Within the sheets [(UO₂)(OH)O₄] pentagonal bipyramids share equatorial edges to form chains, which are cross-linked by [(UO₂)O₄] and [UO₄(H₂O)(OH)] square bipyramids and through hydroxyl groups shared between [(UO₂)(OH)O₄] pentagonal bipyramids. The sheets are pillared by sharing the apical oxygen atoms of the [(UO₂)(OH)O₄] pentagonal bipyramids with the oxygen atoms of [NiO₂(H₂O)₄] octahedral units. That builds a three-dimensional framework with water molecules pointing towards the channels. On heating [Ni(H₂O)₄]₃[U(OH,H₂O)(UO₂)₈O₁₂(OH)₃] decomposes into NiU₃O₁₀.     

Determination of alkali, alkaline earth and transition metal ions in UO2, ThO2 powders and sintered (Th,U)O2 pellets by ion chromatography

An ion chromatographic method has been developed for the determination of traces of Li⁺, Na⁺, K⁺, Ca²⁺, Mg²⁺, Sr²⁺, Fe³⁺, Cu²⁺, Ni^2+, Co²⁺, Zn²⁺, Cd²⁺, Mn²⁺ in UO₂, ThO₂ powders and sintered (Th,U)O₂ pellets. This new method utilizes poly-(butadiene-maleic acid) (PBDMA) coated silica cation exchange column and mixed functionality column of anion and cation exchange to achieve the separation of alkali, alkaline earths and transition metal ions, respectively. It involves matrix separation after sample dissolution by solvent extraction with TBP (tri butyl phosphate)-TOPO (tri octyl phosphine oxide)/CCl₄. Interference of transition metal ions in the determination of alkali, alkaline earth metal ions are removed by using pyridine 2,6-dicarboxylic acid (PDCA) in the tartaric acid mobile phase. Mobile phase composition is optimized for the base line separation of alkali, alkaline earth and transition metal ions. Linear calibration graphs in the range 0.01–20 μg mL⁻¹ were obtained with regression coefficients better than 0.999. The respective relative standard deviations were also determined. Recoveries of the spiked samples are within ±10% of the expected value. The developed method is authenticated by comparison with certified standards of UO₂ and ThO₂ powders.     

Determination of iron and titanium in kaolins by the method of non-dispersive X-ray fluorescence

The possibility of application of the radioactive source excited X-ray fluorescence analysis for titanium and iron determination in kaolins to the routine test of the refinement process has been studied. The iron content can be determined with a simple counting system using a single-channel pulse height analyser, argon filled proportional counter and¹⁰⁹Cd source of 3 mCi for the excitation of K Fe rays. The samples were analysed both as pellets and powders. The iron content ranged from 0.2–2.5% and titanium from 0.1–0.64%. For simultaneous determination of titanium and iron a Si(Li) spectrometer has been used. The²³⁸Pu source has been used for K Fe and K Ti excitation. It is the most convenient source for simultaneous determination of titanium and iron.⁵⁵Fe is the most efficient source for the determination of titanium alone. The best values of precision and determination limit have been achieved for iron with²³⁸Pu and for titanium with⁵⁵Fe.     

Li2O-TiO2-SiO2-P2O5和Li2O-TiO2-Al2O3-P2O5体系锂离子固体电解质的制备及性能研究

一些具有NASICON型网格结构的固体电解质具有高的电导率和好的稳定性,NASICON的意思是Na Super Ionic Conductor[1]。当NaZr2(PO4)3中P5 被Si4 部分取代时便可以得到具有NASICON结构的Na1 xZr2SixP3-xO12体系,其具有高的钠离子电导率。然而有相同结构的Li1 xZr2SixP3-xO12体系的离子电导率却很低,这是因为Li 半径太小,而NASICON三维网格结构的离子通道太大,两者不匹配而使电导率下降[2]。但当LiZr2(PO4)3中Zr4 被离子半径小些的Ti4 取代,所得LiTi2(PO4)3的通道就与Li 半径相匹配,适合于锂离子的迁移,从而使其电导率…     

A multi-method approach for determination of radionuclide distribution in trinitite

The spatial distribution of radiation within trinitite thin sections have been mapped using alpha track radiography and beta autoradiography in combination with optical microscopy and scanning electron microscopy. Alpha and beta maps have identified areas of higher activity, and these are concentrated predominantly within the surficial glassy component of trinitite. Laser ablation-inductively coupled plasma mass spectrometry (LA-ICP-MS) analyses conducted at high spatial resolution yield weighted average ²³⁵U/²³⁸U and ²⁴⁰Pu/²³⁹Pu ratios of 0.00718 ± 0.00018 (2σ) and 0.0208 ± 0.0012 (2σ), respectively, and also reveal the presence of some fission (¹³⁷Cs) and activation products (^152,154Eu). The LA-ICP-MS results indicate positive correlations between Pu ion signal intensities and abundances of Fe, Ca, U and ¹³⁷Cs. These trends suggest that Pu in trinitite is associated with remnants of certain chemical components from the device and surrounding Trinity test-related structures at ground zero. In contrast, negative correlations between Pu ion signals and SiO₂ and K₂O contents were observed within the glassy matrix of trinitite. This LA-ICP-MS result was corroborated by combined back-scattered electron imaging and alpha radiography, and indicates that Pu was not incorporated into unmelted crystalline grains of precursor minerals (i.e., quartz-SiO₂ and K-feldspar-KAlSi₃O₈) present within the desert sand at the Trinity site. The results from this study indicate that the device-related radionuclides were preferentially incorporated into the glassy matrix in trinitite.     

Radiolysis and radiolytic by-product of N,N-diethylhydroxylamine in HNO3 at lower dose

N, N-diethylhydroxylamine (DEHA) is a novel salt-free reducing reagent used in the separation of Pu and Np from U in the treatment of used nuclear fuel. This paper reports on the radiation damage and radiolytic by-product of 0.5 mol L^?1 DEHA in 0.3 mol L^?1?~?1.0 mol L^?1 HNO₃ at dose up to 25 kGy. Results show that the radiolysis rate of DEHA is less than 10%. The main radiolytic products are hydrogen, acetaldehyde, acetic acid and nitrous acid, which increase with the dose. The concentration of acetaldehyde and acetic acid is much higher than that of nitrous acid.

     

Radiation-induced darkening of ionic liquid [C4mim][NTf2] and its decoloration

The radiation effect on a hydrophobic room-temperature ionic liquid (RTIL), 1-butyl-3-methyl-imidazolium bis[(trifluoromethyl)sulfonyl]imide ([C₄mim][NTf₂]), was studied by γ-irradiation under nitrogen atmosphere. Accompanied by color darkening and increase of light absorbance in a wide wavelength range, a distinct absorption peak at around 290 nm for irradiated [C₄mim][NTf₂] appeared when acetonitrile was used as solvent, and the intensity of the peak enhanced with increasing dose. The spectrophotometric study on the irradiated RTILs containing 1,3-dialkylimidazolium cations associated with different inorganic anions revealed that the peak is ascribed to the radiolysis products of the [C₄mim]⁺. And the wavelength of the peak was affected by alkyl chain length on imidazolium cation, while the intensity of the peak was influenced by anions. With incorporating a little amounts of oxidants, such as KMnO₄ and HNO₃ into irradiated [C₄mim][NTf₂], the intensity of the peak at 290 nm decreased obviously and the decoloration of [C₄mim][NTf₂] occurred, suggesting that the peak at 290 nm is assigned to the colored species and the species can be oxidized.     

Coordination environment of [UO2Br4] in ionic liquids and crystal structure of [Bmim]2[UO2Br4]

The complex formed by the reaction of the uranyl ion, UO₂²⁺, with bromide ions in the ionic liquids 1-butyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([Bmim][Tf₂N]) and methyl-tributylammonium bis(trifluoromethylsulfonyl)imide ([MeBu₃N][Tf₂N]) has been investigated by UV–Vis and U L_III-edge EXAFS spectroscopy and compared to the crystal structure of [Bmim]₂[UO₂Br₄]. The solid state reveals a classical tetragonal bipyramid geometry for [UO₂Br₄]²⁻ with hydrogen bonds between the Bmim⁺ and the coordinated bromides. The UV–Vis spectroscopy reveals the quantitative formation of [UO₂Br₄]²⁻ when a stoichiometric amount of bromide ions is added to UO₂(CF₃SO₃)₂ in both Tf₂N-based ionic liquids. The absorption spectrum also suggests a D_4h symmetry for [UO₂Br₄]²⁻ in ionic liquids, as previously observed for the [UO₂Cl₄]²⁻ congener. EXAFS analysis supports this conclusion and demonstrates that the [UO₂Br₄]²⁻ coordination polyhedron is maintained in the ionic liquids without any coordinating solvent or water molecules. The mean U–O and U–Br distances in the solutions, determined by EXAFS, are, respectively, 1.766(2) and 2.821(2) Å in [Bmim][Tf₂N], and, respectively, 1.768(2) and 2.827(2) Å, in [MeBu₃N][Tf₂N]. Similar results are obtained in both ionic liquids indicating no significant influence of the ionic liquid cation either on the complexation reaction or on the structure of the uranyl species.     

Uranyl-catalyzed chemiluminescent reaction of U4+ oxidation by dioxygen in aqueous HClO4 solution

Chemiluminescence (CL) accompanying the reaction of U⁴⁺ with O₂ in 0.0004–0.1M HClO₄ was studied. It was found that the electron-excited uranyl ion (UO₂ ²⁺)^* is the CL emitter. The fact that the reaction rate and the CL yield increase as the solution acidity decreases was explained by different reactivities of the U _aq ⁴⁺ aquation and the products of its stepwise hydrolysis, UOH³⁺ and U(OH)₂ ²⁺, toward O₂. Based on the results of analysis of the chain-radical mechanism of the reaction between U⁴⁺ and O₂, it was concluded that transfer of an electron from the UO₂ ⁺ ion to the oxidizing agent (a ·OH radical) is the most plausible elementary step of the reaction of (UO₂ ²⁺)^* formation. It was found that the reaction rate, as well as the CL yield, increase substantially in the presence of uranyl ion. Catalytic action of UO₂ ²⁺ was explained by the formation of a UO₂ ²⁺·UO₂ ⁺ complex, which reduces the rate of the UO₂ ⁺ disproportionation reaction (UO₂ ⁺ is an intermediate of the reaction and is involved in chain propagation), and by regeneration of the active center, UO₂ ⁺, in the reaction of UO₂ ²⁺ with U⁴⁺. Published inIzvestiya Akademii Nauk. Seriya Khimicheskaya, No. 9, pp. 1522–1528, September, 2000.