[en] Highlights: • The Artemisinin dimorphic system was found to be enantiotropic. • The Orthorhombic modification is the form stable at low-temperatures and the triclinic modification the form stable at high-temperatures. • The polymorphic phase transition occurs at ∼130 °C. - Abstract: The polymorphism of the anti-malarial compound artemisinin was examined. The phase behavior of solid artemisinin has experimentally been investigated using differential scanning calorimetry and temperature-resolved X-Ray powder diffraction. In addition, complementary solution studies and suspension experiments were performed. The results clearly confirm the existence of two modifications of artemisinin, which are related enantiotropically. The orthorhombic modification is the thermodynamically stable form at low temperatures, while the triclinic form is the stable one at higher temperatures with a transition temperature of ∼130 °C. Problems associated with analysis of the polymorphic phase behavior are comprehensively addressed

[en] Manganese dioxide is present in some of the radioactive waste storage tanks within the U.S. Department of Energy complex and may influence the oxidation state of plutonium in these waste solutions. Manganese dioxide also is a common component of natural water systems in which it may influence the oxidation state and subsequently the environmental behavior of plutonium. Because of its oxidizing properties, manganese dioxide has been considered in a sediment remobilization model as contributing to the oxidation of sorbed Pu(IV) to the more soluble Pu(V).1 In this study an experimental procedure has been developed to monitor the oxidation kinetics of plutonium by manganese dioxide

[en] The hypochlorite ion (OCl-), a major product of the radiolysis of water in solutions containing high concentrations of chloride, can be expected to influence the oxidation state distribution of plutonium in these solutions. Since plutonium has significantly different chemistry in each of its oxidation states, knowledge of the oxidation state distribution of plutonium and of the kinetics of transfer between these oxidation states is essential for modeling the behavior of plutonium in aqueous systems and for design of efficient remediation procedures for plutonium containing wastes

[en] The kinetics of the reduction of Pu(V)O₂⁺ by hydrogen peroxide have been studied in 1.0 m NaCl solutions at 19±2 C. Experiments were conducted over a range of pH values from 7.9 to 10.8 with variable concentrations of hydrogen peroxide between 1.00.10^-5 M and 4.00.10² M. The concentration of plutonium in solution was (1.5±0.2) . 10^-10 M. The decrease of the concentration of PuO₂⁺ in solution with time was monitored by ultrafiltration combined with solvent extraction using TTA. Independent extraction experiments, using DBM as extractant, were employed to validate the results. The reduction of PuO₂ was found to be first order with respect to hydrogen peroxide concentration and inverse first order with respect to hydrogen ion concentration. With these results and assuming a first order dependence with respect to the concentration of PuO₂^-, the reaction can be described by the following rate equation: d[Pu(V)]/dt = k . [Pu(V)] . [H₂O₂]/[H^-], with k = 3.59 . 10^-9 ± 1.79 . 10^-9 min^-1. Using this equation, the redox half life of PuO₂^- has been calculated for a range of environmentally relevant conditions. (orig.)

[en] The kinetics of the oxidation of plutonium(IV) by manganese dioxide were studied in 1.0 M NaCl over the pH range from 2.5 to 8.2 with variable concentrations of manganese dioxide from 0.01 m²/L to 4.97 m²/L at a constant temperature of 19±2 C. The concentration of plutonium in solution was 6.0(±1.0) x 10^-10 M. Fractions of Pu(IV), (V) and (VI) as a function of time were determined by removal of plutonium from the solid phase followed by an ultrafiltration/solvent extraction procedure using TTA and HDEHP as extractants. Appropriate removal conditions were established for Pu(IV), Pu(V) and Pu(VI) using Th(IV), Np(V) and U(VI) as oxidation state analogs. In the pH range from 2.0 to 3.5, the oxidation of Pu(IV) by manganese dioxide was first order with respect to the concentration of manganese dioxide and -0.21 with respect to the hydrogen ion concentration. Consequently, assuming a first order dependence with respect to the concentration of Pu(IV), the oxidation reaction can be described by the following rate equation: -d[Pu(IV)]/dt = k . [Pu(IV)] . [MnO_2(s)] . [H⁺]^-0.21 with k = 3.72(±0.13) x 10^-3 (m^-2 L)(mol^-1 L)^-0.21(min)^-1. Using the kinetic data determined in this study, the influence of manganese dioxide on the oxidation state distribution of plutonium under various environmental conditions as well as in waste solutions can be estimated. (orig.)

[en] Radionuclide separation processes that have been used successfully in the past as well as recent developments in separation techniques are discussed with emphasis on their applications in the treatment of radioactive wastes. Both aqueous and nonaqueous processes are evaluated. Separation procedures proposed for use in systems involving transmutation of nuclear waste products are also discussed. (author)

[en] Calorimetry is widely used in US nuclear facilities as a prime non-destructive assay technique for the accountancy of weapons grade plutonium materials. Ample experiences and performance data from this application do exist, underlining the high measurement accuracy achievable with this measurement technique for the plutonium assay of low-burnup materials. The non-destructive assay approach for this type of plutonium benefits from the fact that the relative isotope abundance of the main plutonium isotopes responsible for the heat production, namely ²³⁹Pu and 240 Pu, can be determined fairly well by gamma spectrometry using the common isotopic analysis code MGA (and by mass spectrometry anyway). By contrast, only limited experimental performance data are so far available for the calorimetric assay of reactor-grade plutonium materials. The accuracy limits of the technique for higher burnup plutonium will be primarily not set by the heat measurement itself, but by the ability to determine accurately the relative abundance of the main heat producing minor isotopes 238 Pu and ²⁴¹Am, which are typically responsible for 60-80 % of the thermal power generated by reactor-grade plutonium. We have carried out an experimental performance study for the calorimetric assay of reactor-grade plutonium, based on measurements on a series of PuO₂ powder and MOX pellet samples. The amount of plutonium in the samples was in the range from 0.2-0.8 g, yielding thermal power values between about 2-12 mW. The heat measurements were carried out in two different small sample calorimeters, both using highly sensitive thermopile sensors for the heat flow measurements. The extremely low noise of this type of heat flow sensors allows for high-precision measurements of low-power items, reaching a precision and accuracy of the order of 0.1 % at a thermal power level of 10 mW. The necessary plutonium and ²⁴¹Am isotope abundances needed for the conversion of the measured thermal power into the corresponding amount of plutonium were made both by gamma spectrometry using the analysis code MGA (yielding the Pu and ²⁴¹Am abundances), and by mass/alpha spectrometry (for the Pu isotope abundances). Special efforts were undertaken to determine accurate reference values for the isotope abundance of 238 Pu, which contributed the largest portion to the thermal power of the measured samples. In routine Safeguards measurements only moderate accuracy requirements are normally imposed to the determination of this minor isotope (international target values quote random and systematic uncertainties of 1.5% and 1%, respectively for high burnup plutonium). However, these levels of uncertainty are far from being adequate for an accurate assay of reactor-grade plutonium by calorimetry. For this reason efforts were made to achieve, besides the routine analysis by means of alpha- and gamma-spectrometry, also a reliable 238 Pu abundance measurement by exclusively using mass spectrometry. This necessitated a careful U-Pu separation in order to avoid the 238 Pu-²³⁸U isobaric interference in mass spectrometry. Separation factors exceeding 10 6 were achieved by means of a two- fold extraction chromatographic separation using Eichrome's TEVA resin. The accurate 238 Pu abundance values determined by mass spectrometry from the separated plutonium fraction not only helped for the interpretation of the calorimeter measurements, but also served for the validation of the 238 Pu results obtained by gamma spectrometry. The plutonium assay results derived from the calorimeter measurements with isotopic input data from gamma and mass spectrometry are compared and evaluated relative to reference values obtained by a classical titration analysis. Performance values for the calorimetric plutonium assay are presented and discussed

[en] An experimental procedure has been developed for the analysis of neptunium in input solutions of spent nuclear fuel reprocessing. The method consists of selective chromatographic extraction of neptunium and plutonium using TEVA resin followed by determination of the ratio of [Np]/[Pu] on the column by high-resolution gamma spectrometry. Knowing the concentration of plutonium from routinely performed K-edge densitometry and X-ray fluorescence measurements, the concentration of Np can be evaluated. The method has been tested on simulated solutions and validated on spent fuel of known [Np]/[Pu] ratio. (orig.)

[en] In this work we focused on the age determination of uranium materials of different uranium enrichment. The radioactive decay of the uranium isotopes provides a chronometer that is inherent to the material, in particular the mother/daughter pairs ²³⁴U/²³⁰Th and ²³⁵U/²³¹Pa can be advantageously used. Due to the relatively long half-lives of ²³⁴U (2.46 · 10⁵ years) and ²³⁵U (7.04 · 10⁸ years) only minute amounts of daughter nuclides are growing in, therefore both separation of Th and Pa from uranium must be of high chemical recovery and must afford large decontamination factors. Analytical methods for the age determination of uranium samples using the parent/daughter relations ²³⁴U/²³⁰Th and ²³⁵U/²³¹Pa is demonstrated. Thorium is separated from bulk uranium using extraction chromatography and subsequently quantified using square-spectrometry, thermal ionisation mass spectrometry (TIMS) and inductively coupled mass spectrometry (ICP-MS). Protactinium is separated by highly selective sorption of protactinium to silica gel followed by square-spectrometric quantification. The methods were tested and validated using uranium reference materials of different uranium enrichment and of known ages. The experimental results obtained with both methods were found to agree with the assumed ages of the reference materials within the combined uncertainty of the measurement. The analysis exploiting the parent/daughter pair ²³⁵U/²³¹Pa exhibits a slightly larger combined uncertainty and bias than the thorium method but is found valuable in validating the experimental results by means of a second, independent analysis

[en] This paper describes the analytical methods (thermal ionization mass spectrometry, inductively coupled plasma mass spectrometry, and alpha spectrometry) that have been developed for determination of the age of uranium and discusses their advantages and limitations. With regard to potential application of the methods (e.g. Fissile Material Cut-off Treaty), the discussion focuses on highly enriched uranium, because this seems to be of highest strategic relevance. The different analytical methods were tested and validated by use of uranium reference materials of different ²³⁵U isotope abundance and of known ages. The results show that thermal ionization mass spectrometry and alpha spectrometry are both very accurate and precise techniques for this application. Inductively coupled plasma mass spectrometry, on the other hand, although less precise, because of the different approach to the analytical problem, is still sufficiently accurate to be used as a rapid screening method. (orig.)