Actinide Content Research Articles

Distribution and origin of major and trace elements (particularly REE, U and Th) into labile and residual phases in an acid soil profile (Vosges Mountains, France)

A 7 step sequential extraction procedure has been conducted on a podzolic soil profile from the Vosges Mountains in order to determine the ability of several elements to be released to the environment. Very little Si, K and Al were extractable (<10% of the total soil concentration) but larger proportions (> 10% of the total soil concentration) of Ca, P, metals (Fe, Pb), REE and actinides (Th, U) could be leached. For each element, preferential binding sites can be recognized. High recovery of P and Ca in the acid soluble fraction (AS) suggests that phosphate minerals are highly involved in this step of the extraction. Organic matter appears to control the adsorption of Ca, Fe, Th, U and REE, even at depths in the soil profile where organic matter content is particularly low (0.5%). Weak acid leaching experiments (with HCl and acetic acid 0.001 N) were also performed in order to characterize the origin of the insoluble material in this soil profile. The leachable REE distributions indicate that a large part of the labile REE in the surface horizon has an atmospheric origin whereas at greater depth phosphate mineral (apatite) alteration is the main factor controlling REE release in the leachate. The study further suggests that adsorbed material holding actinides and REE are not strictly the same. So, caution should be taken when using REE as analogues for actinides in soils systems.

Effect of Alpha Decay on Nuclear Borosilicate Glass Properties

AbstractThe properties of actinide glasses are studied in the context of high-level waste management programs. Reprocessing high burnup fuels in particular will increase the minor actinide content in the glass package, resulting in higher cumulative alpha decay doses in the glass, and raising the question of the glass matrix behavior and especially its containment properties. The effect of alpha self-irradiation on the glass behavior is evaluated by doping the glass with a short-lived actinide (244Cm) to reach in several years the alpha dose received by the future glass packages over several thousand years. “R7T7” borosilicate glasses were doped with 3 different curium contents (0.04 wt%, 0.4 wt% and 1.2 wt% 244CmO2). The glass homogeneity was characterized by SEM and XRD. Chemical analysis after dissolution of the glass confirmed the chemical composition. The curium content and its distribution in the glass specimens were checked by calorimetry measurements and gamma spectrometry on 243Cm and 245Cm isotopes. The density and mechanical properties of the curium-doped glasses were characterized up to 2 × 1018 α/g, revealing only a slight evolution of the macroscopic behavior of R7T7 glass in this range. The leaching behavior of curium-doped glass was also studied by Soxhlet tests. The results do not show any significant evolution of the initial alteration rate with the alpha dose.

Evaluation of a single collector, double focusing sector field inductively coupled plasma mass spectrometer for the determination of U and Pu concentrations and isotopic compositions at trace level

This work explores the performance of the “Axiom”, a double focusing sector field ICP-MS (ICP-SFMS) in the determination of actinide concentration and isotopic ratio at trace level. On the actinide mass range the performances observed are characterized by high sensitivity, around 2.8·106 cps·µg−1·l, and low background, below 0.3 cps. Therefore, the absolute instrumental detection limit is approximately 0.05 fg for Pu isotopes. Furthermore, the 235U/238U ratio for a 0.5 µg·l−1 U500 isotopic standard could now be measured using the ICP-SFMS with a relative standard deviation less than 0.1%. Moreover, the accuracy of the measured ratio was demonstrated at low concentration with the target value remaining within experimental uncertainty limits.

Statistical analysis of measured actinide concentration bias in a Mixed Analyte Performance Evaluation Program

Data from seven Mixed Analyte Performance Evaluation Program studies have been used to identify specific problems with results and/or laboratories. These data have never been used to determine the bias or variability in data reporting for these laboratories in general. These historical analytical data were analyzed for accuracy and precision overall and separated by method of sample preparation, method of sample pretreatment, method of detection and sample size. The calculated mean percent relative bias was analyzed for the actinides (241Am, 238Pu, 239/240Pu, 234/233U, and 238U) in the soil and water standards. Results from most analysis methods were at least slightly negatively biased. For the water standards, the overall 95% confidence interval for the mean percent relative bias was −1.6% to −0.5% indicating that the average actinide results slightly underestimated the true concentration. For the soil standards, the overall 95% confidence interval for the mean percent relative bias was −8.1% to −6.6% indicating that the average actinide results moderately underestimated the true concentration.

Concentration and Separation of Actinides in Nitric Acid Media with Fibrous “Filled” Sorbents

The behavior of actinides in sorption from nitric acid solutions with fibrous “filled” sorbents was studied. It was found that the sorbents with methylpyrazole groups POLIORGS 4-n and 17-n and anion exchanger AV-17-n sorb completely only Pu(IV) from strongly acidic solutions. The other actinides studied [Am(III), U(VI), Pa(V), Np(V)], and also Eu(III), Tc(VII), and Cs(I) are not appreciably sorbed under these conditions. The sorbents are characterized by good kinetic properties and chemical stability in HNO3 solutions. The possibility of using fibrous "filled" sorbents for concentrating plutonium from nitric acid solutions and separating plutonium from other radionuclides was demonstrated. Pu can be desorbed with 0.5-1 M HNO3.

A comparative physics study of alternative long-term strategies for closure of the nuclear fuel cycle

The appropriate management of radioactive waste arising from the nuclear fuel cycle is considered to be a key issue in the development of future, more sustainable nuclear energy systems. In this context, the partitioning and transmutation of actinides could play an important role through the achievement of very significant reductions in the actinide content and radiotoxicity of the high-level waste requiring geological disposal. The current paper reports on the results of a detailed physics study carried out to compare the pros and cons of alternative strategies for closure of the nuclear fuel cycle. Different long-term “steady-state” scenarios have been considered, involving the deployment, to varying degrees, of light water reactors (LWRs) and advanced fast-spectrum systems. The same nuclear data and calculation methods have been used throughout, so that a consistent and reliable comparison of the relative performance of the three basic fuel cycle options (once-through, plutonium recycle, and recycling of all actinides) has been made possible. In addition, with transmutation having been considered employing both critical and accelerator-driven fast-spectrum systems, the study has provided an evaluation of the advantages and disadvantages of these two different advanced system types.

Considerations for Fusion-Based Actinide Destruction

A fusion based actinide destruction system is advantageous because of higher actinide destruction efficiency and higher energy efficiency when compared to other destruction technologies. The unique neutron multiplication capability due to the n,2n reactions in blanket materials with 14 MeV D-T neutrons enhances further the performance efficiency.Investigation of a high performance fusion based actinide destruction system was conducted. A self-cooled, actinide-carrying molten salt blanket can be designed to operate at a high sub-criticality factor of 0.95-0.96, with less than 0.4 wt% actinide concentration dissolved in the molten salt. The corresponding blanket energy multiplication is 160. Lithium-6, which is required for tritium breeding, can be used as a variable to shape the neutron spectrum and control the criticality factor, and thus to maintain a constant fission thermal power output from the actinide destruction plant.Sub-criticality can be maintained in all cases of the actinide destruction plant, during normal operation and abnormal conditions.A fusion device projected from a tokamak experiment can produce 30 MW fusion power, with a plasma amplification factor of 2. It is considered adequate to drive the sub-critical molten salt blanket. The total thermal fission power is about 4000 MW, which is able to destroy 1.6 metric tons of actinides annually when operating at full power.

Estimation of actinide skeletal content in humans based on bone samples collected at autopsy.

The USTUR has developed simple linear and multiple regression models for estimating skeletal actinide concentrations on the basis of bone samples collected at autopsies of non-whole body tissue donors. Bone samples usually collected include a clavicle, the patella(e), one or more ribs, the sternum, and a vertebral wedge cut from within the abdominal cavity. The described models were derived by regression analyses with the analytical results from those bones and the entire skeletons of eight whole body donations to the USTUR. With the model, skeletal concentrations of 238Pu, (239+240)Pu, and 241Am can be estimated from wet or ashed actinide concentrations in one to five of the bones usually collected at autopsy and analyzed. Application of the models to a selected USTUR non-whole body donation (Case 0240) indicated that the skeletal actinide concentration estimates were reasonably precise and that there was good agreement between the results from individual bones with wet or ashed actinide concentrations. The USTUR will apply the model that is based on wet concentrations of bones to estimate skeletal concentrations of actinides in all non-whole body autopsy cases for the sake of consistency because of the large number of early cases for which ashed weights of bones were not recorded.

Effect of Synthesis Conditions and Actinide Contents on Phase Composition of Actinide Bearing Ceramics

ABSTRACTPu-bearing zirconolite and pyrochlore based ceramics were prepared by melting under oxidizing and reducing conditions at 1550 °C. 239Pu content in the samples ranged between ∼10 and ∼50 wt.%. Phase composition of the ceramics and Pu partitioning were studied using X-ray diffraction (XRD) and scanning electron microscopy with energy dispersive system (SEM/EDS). Major phases in the samples were found to be the target zirconolite and pyrochlore as well as a cubic fluorite structure oxide. Normally the Pu content in the Pu host phases was 10–12 wt.%. This corresponds to the Pu content recommended for matrices for immobilization of excess weapons plutonium. At higher Pu content (up to 50 wt.%) additional phases, such as a PuO2-based cubic fluorite-structured solid solution, perovskite, and rutile were found.

Actinide Speciation in Environmental Remediation

Environmental actinide remediation is often performed in the absence of a clear understanding of the contaminant chemical forms. Remediation efforts are more concerned with the initial and final concentration of the contaminant actinides. Allowable final concentrations of actinides in remediation sites are mainly based on expected dose or prenegotiated levels. However, understanding the chemical forms of actinides in the environment is key to assessing their long-term behavior and developing enhanced analytical methods for their environmental detection. The dominant environmental chemical form will dictate the actinide transport behavior through solubility, colloid formation, or the interaction with natural ligands/ This information can be used to evaluate the fate and potential transport of actinides. If the dominant environmental chemical species are understood, analytical methods, for enhanced counting statistics and a reduction of counting error can be developed. A reduction in counting error will increase the fraction of verifiable waste below a given threshold, thereby increasing the rate of site clean up while reducing the cost. The impact of actinide speciation is presented in 2 sites contaminated with Am and Pu undergoing remediation. Actinide speciation has affected remediation strategies, regulatory response, and costs.

Predicting and understanding the behaviour of minor actinide targets for transmutation

The irradiation behaviour of minor actinide targets containing high concentrations of minor actinides (MA) is not yet well understood, partly because of a lack of such experiments and partly due to the fact that their physical and thermodynamic properties are not well known. By a systematic review and analysis of the literature, the required quantities have been extrapolated/approximated and the expected thermal behaviour is described. In addition the chemical changes in the fuel, resulting from the 'chemical revolution' taking place in the transmutation process, are described by chemical thermodynamic techniques. Based on these calculations, different target concepts such as (U,MA)O2, (Zr,Y,MA)O2, Mo+(Zr,Y,MA)O2 and MgO+(Zr,Y,MA)O2 are discussed.

Simulation of Mixed-Oxide and Low-Enriched Uranium Fuel Burnup in a Pressurized Water Reactor and Validation Against Destructive Analysis Results

This work examines the capabilities of simulation codes to predict the concentration of nuclides in spent reactor fuel, in particular mixed-oxide (MOX) fuel, via comparisons with destructive radiochemical analyses performed on irradiated samples. We report on three MOX samples irradiated in a pressurized water reactor (PWR) and two UO2 samples irradiated in a different PWR. Actinide and fission-product concentrations were measured and were compared with concentration values obtained from simulation studies. The actinides include isotopes of uranium, neptunium, plutonium, americium, and curium. The fission products include isotopes of cesium, neodymium, samarium, europium, and gadolinium as well as 90Sr, 95Mo, 99Tc, 101Ru, 106Ru, 103Rh, 109Ag, 125Sb, 129I, and 144Ce. For many of the actinides, the predictions are quite good when compared with the measured values; but concentrations of some tend to be overpredicted. The cesium and neodymium, and some samarium concentrations, are well predicted, but some of the other fission products show variable results. The sensitivity of some of the results to sample-burnup estimates is discussed.

Advantages of Irradiated DUPIC Fuels from the Perspective of Environmental Impact

This study compares some properties of irradiated Direct Use of Spent Pressurized Water Reactor (PWR) Fuel In Canada Deuterium Uranium reactor (CANDU) (DUPIC) fuels with properties of other fuel cycles. The properties include the radiotoxicity, decay heat, activity, and actinide content embedded in various spent fuels or high-level wastes, which could be measures of the effectiveness of waste management. From radiotoxicity analysis of fuel cycles, the toxicity of the DUPIC option based on 1 GW(electric)·yr is much smaller than those of other fuel cycle options such as the PWR once-through mode, mixed oxide fuel recycling mode, and CANDU once-through mode. The analysis shows that the value is just about half the order of magnitude of other fuel cycles until decayed to a level below the toxicity of initial ore. This means that the DUPIC option could have an indirect benefit on the environmental effects of long-term spent-fuel disposal. From total activity analysis of various fuel cycle options, the activity per metric ton heavy metal of spent fuel is the lowest in natural uranium CANDU fuel, but in the case of activity based on 1 GW(electric)·yr, the DUPIC option has the smallest activity. In the meanwhile, from the activity analysis of 99Tc and 237Np, which are important to the long-term transport in geologic media, the DUPIC option was being contained in only about half of those other options. In conclusion, compared to other fuel cycle cases, the irradiated DUPIC fuels would have good properties from the perspective of environmental effects.

Ion irradiation-induced amorphization and nano-crystal formation in garnets

The radiation susceptibility of garnet structure types (A 3B 2(XO 4) 3, I a3d, Z=8) has been examined by 1.0 MeV Kr 2+ irradiation with in situ transmission electron microscopy over the temperature range of 50–1070 K. The target garnets included five natural garnets: almandine, andradite, pyrope, grossular and spessartine, and five synthetic garnets incorporating various contents of actinides. The synthetic garnets were silicates (N-series) and ferrate–aluminates (G-series). The critical amorphization temperature ( T c), above which amorphization does not occur, were determined to be 1050 K for N77, 1130 K for N56, 1100 K for G3, 890 K for G4 and 1030 K for andradite. T c of the synthetic garnets increased with increasing mass density of the target and as the electronic-to-nuclear stopping power ratio decreased. During ion irradiation of the G3 garnet at a temperature of 1023 K near the T c, nano-crystals were produced in which the unit-cell parameter of the original garnet was halved (∼0.64 nm) as a result of a radiation-induced recrystallization process, without evidence for ordering of the cations.

Geochemical Characteristics and Variations of the Ryoke Granitoids, Southwest Japan: Petrogenetic Implications for the Plutonic Rocks of a Magmatic Arc

Geochemical characteristics and variations of the Ryoke granitoids, southwest Japan, have been elucidated, based on more than 70 analyses from the Mikawa and Kinki districts for major and trace elements including REE. The Ryoke granitoids are the products of a continental-margin arc magmatism of Pacific Asia during late-Cretaceous period, and represent its magmatic front. They are classified into the Older and Younger groups, according to their intrusive ages and textural and petrographic characteristics. The Older Ryoke granitoids were emplaced at ca. 20 km depth during 100-85 Ma, and deformed and recrystallized together with the host Ryoke metamorphic rocks. The Younger Ryoke granitoids are post-tectonic plutons, emplaced at ca. 15 km depth (85-75 Ma), and accompanied by the formation of contact aureoles. The majority belongs to I-type and ilmenite-series granitoids. The granitoids of both groups are geochemically quite similar: calcic (alkali-lime index ∼65), very low in Fe +3/Fe +2 ratio and both metaluminous and peraluminous. They have low Rb (<100 ppm) and actinide contents, and moderate abundances of other LIL and HFS elements. They fall exclusively within the volcanic arc granite field in all the trace element discrimination schemes. Trace element geochemistry and chondrite-normalized REE patterns suggest that the I-type Ryoke granitoids were generated by the dehydration melting of amphibolite, of basaltic and basaltic-andesitic composition, or hydrous melting of tholeiitic basalt at 1 GPa and higher pressures. The heat-source for the melting of protoliths should be the underplating basalts, of which remnants are seen as the mafic rocks enclosed and/or intruded by the Ryoke granitoids.

Competition between disorder creation and annealing in fluoroapatite nuclear waste forms

He-ion irradiation simulating α-particle emission, induces crystal defect recovery in fluoroapatite Ca 10(PO 4) 6F 2 pre-damaged by Pb-ions. In fluoroapatite loaded with α-emitter actinides, a similar effect is expected. Therefore, the amorphization level in fluoroapatite has been modeled versus disposal time, α-decay fluence and temperature. When loaded with 239 Pu and 244 Cm , the structural disorder reaches a plateau, which level depends only slightly on α-energies. This level corresponds to an amorphous fraction of approximately 20% of disorder and is independent of the dose rate. The time delay in reaching the equilibrium regime in the disorder production is inversely proportional to the dose rate whereas the dose or the actinide content is not a critical parameter for the magnitude of the equilibrium level.

Ion Irradiation Effects in Synthetic Garnets Incorporating Actinides

ABSTRACTRadiation durability of garnet [A3B2(XO4)3; Ia3d; Z=8] has been examined by 1.0 MeV Kr2+ irradiation with in situ transmission electron microscopy over the temperature range of 50 to 1070 K. The targets were five synthetic garnets incorporating various contents of actinides and andradite, Ca3Fe2Si3O12. The synthetic garnets were silicates (N series) and ferrate-aluminate series (G series).The critical amorphization temperatures (Tc), above which amorphization does not occur, were determined to be 1050 K for N77, 1130 K for N56, 1100 K for G3, 890 K for G4 and 1030 K for andradite. Tc of the synthetic garnets increased as the average atomic mass of the garnet increased. The maximum transferred energy by ballistic interaction was positively correlated to the atomic mass. The larger cascade size that formed due to the larger Emax might lead to the higher Tc.

Natural Ground Water Colloids From the USGS J-13 Well in Nye County, NV: a Study Using SAXS and TEM

We report results from ultra small angle x-ray scattering (USAXS) and transmission electron microscopy (TEM) of dilute silicate colloids that occur naturally in groundwater from the USGS J-13 well, located near the Yucca Mountain Site in Nevada. We also examined a separate sample of this groundwater that had been treated by heating to 90 degrees C in contact with crushed Topopah Spring Tuff from the Yucca Mountain. The USAXS measurements were done at the UNICAT undulator beamline at the Advanced Photon Source at the Argonne National Laboratory. Power-law plots (scattering intensity verses momentum transfer) were fitted to the USAXS data. Colloids in the untreated J-13 groundwater were shown to have a fractal dimension of nearly 3, whereas colloids in the treated groundwater (“EJ-13”) have a dimensionality of approximately 2.4 over a length scale of approximately 3 to 300 nm. Similar power-law plots with dimension 3 characterized concurrent SAXS measurements from aqueous suspensions of Na-montmorillonite and NIST Brick Clay (NBS-67). We attribute these results to the sheet silicate layered structure of the clay colloids present in J-13 well water, montmorillonite, and “brick clay” systems. The differences between EJ-13 and as-received J-13 are perhaps owing to exchange of calcium for sodium with the tuff. Radionuclide incorporation into, adsorption onto, or ion exchange with existing groundwater colloids may promote colloidal transport of radionuclides in groundwater. Such radionuclide-bearing colloids could thereby increase the concentrations of actinides in groundwater and enhance migration into human-accessible aquifers. Our results demonstrate the first application of USAXS to study the physical nature of such groundwater colloids, and represent perhaps one of the most dilute systems ever studied by small angle scatering.

Hydration of Magnesium Oxide in the Waste Isolation Pilot Plant

ABSTRACTMagnesium oxide (MgO) is the only engineered barrier being emplaced in the Waste Isolation Pilot Plant (WIPP), a U.S. Department of Energy repository for transuranic waste. MgO will lower dissolved concentrations of actinides by consuming CO2 from possible microbial activity, by buffering the pH between 8.5 and 9.5, and by reducing the amount of free water in the repository. This paper discusses results from experiments measuring the hydration of MgO. Results suggest that periclase (MgO) hydrates rapidly to brucite (Mg(OH)2) in de-ionized water and 4 M NaCl solution at 90°C; the hydration rate decreases as temperature decreases. In ERDA-6, a NaCl-rich WIPP brine, MgO hydrates directly to brucite; in GWB, a high-Mg brine, periclase hydrates to magnesium chloride hydroxide hydrate(s) until the dissolved Mg2+ concentration decreases, and brucite becomes the stable phase. Under humid conditions MgO fully hydrates at higher humidities(&gt; 50%). All data are consistent with diffusion-limited hydration reactions.

Host Phases for Actinide Elements in the Metallic Waste Form

ABSTRACTArgonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists of stainless steel cladding hulls that contain undissolved metal fission products such as Tc, Ru, Rh, Pd, and Ag; a small amount of undissolved actinides (U, Np, Pu) also remains with the hulls. These wastes will be immobilized in a waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). Scanning electron microscope (SEM) observations of simulated metal waste forms (SS-15Zr with up to 11 wt% actinides) show eutectic intergrowths of Fe-Zr-Cr-Ni intermetallic phases with steels. The actinide elements are almost entirely in the intermetallics, where they occur in concentrations ranging from 1–20 at%. Neutron- and electron-diffraction studies of the simulated waste forms show materials with structures similar to those of Fe2Zr and Fe23Zr6.Dissolution experiments on simulated waste forms show that normalized release rates of U, Np, and Pu differ from each other and from release rates of other elements in the sample, and that release rates for U exceed those for any other element (including Fe). This paper uses transmission electron microscope (TEM) observations and results from energy-dispersive X-ray spectroscopy (EDX) and selected-area electron-diffraction (SAED) to characterize relationships between structural and chemical data and understand possible reasons for the observed dissolution behavior.Transmission electron microscope observations of simulated waste form samples with compositions SS-15Zr-2Np, SS-15Zr-5U, SS-15Zr-11U-0.6Rh-0.3Tc-0.2Pd, and SS-15Zr-10Pu suggest that the major actinide-bearing phase in all of the samples has a structure similar to that of the C15 (cubic, MgCu2-type) polymorph of Fe2Zr, and that materials with this structure exhibit significant variability in chemical compositions. Material whose structure is similar to that of the C36 (dihexagonal, MgNi2-type) polymorph of Fe2Zr was also observed, and it exhibits less chemical variability than that displayed by material with the C15 structure. The TEM data also demonstrate a range of actinide concentrations in materials with the Fe23Zr6 (cubic, Mn23Th6-type) structure.Microstructures similar to those produced during experimental deformation of Fe-10 at% Zr alloys were observed in intermetallic materials in all of the simulated waste form samples. Stacking faults and associated dislocations are common in samples with U, but rarely observed in those with Np and Pu, while twins occurred in all samples. The observed differences in dissolution behavior between samples with different actinides may be related to increased defect-assisted dissolution in samples with U.