Partitioning Of Plutonium Research Articles

Partitioning of iron and plutonium to exopolymeric substances and intracellular biopolymers: A comparison study between the coccolithophore Emiliania huxleyi and the diatom Skeletonema costatum

Iron (Fe), a micronutrient for algal growth, and plutonium (Pu), an anthropogenic radionuclide, share some common features. This includes similar oceanic distributions when different input modes are taken into account, as well as their chemical behavior, such as a high affinity to natural organic matter (NOM). The NOM produced by various phytoplankton communities can potentially influence Fe cycling in the ocean, and likely also influence the transport behavior of Pu. We conducted laboratory incubation experiments using the coccolithophore Emiliania huxleyi and the diatom Skeletonema costatum, in the presence of 59Fe and 238Pu as radiotracers, in order to differentiate Fe and Pu uptake by extracellular exopolymeric substances (EPS) and intracellular biopolymers. The Fe and Pu distributions in select organic compound classes produced by these two types of phytoplankton, including proteins, total carbohydrates (TCHO) and uronic acids (URA), were compared. Our results indicated that most of the Fe and Pu (>95%) were found concentrated in E. huxleyi-derived non-attached EPS, while much less (<2%) was present in the intracellular fraction of E. huxleyi. In contrast, in the diatom S. costatum, Fe and Pu were both distributed with EPS > intracellular biopolymers > outer cell covering (i.e., frustule). In fact, over 50% of the Fe was concentrated in S. costatum-derived attached EPS and intracellular biopolymers. The diatom derived Fe-EPS complexes were more hydrophobic, with stronger tendency to aggregate in seawater. Fe binding to biopolymers in both E. huxleyi and S. costatum cultures was related to URA concentrations, but the overall distribution of URA between these two phytoplankton species was different (e.g., high intracellular abundance of URA in S. costatum but low intracellular URA abundance in E. huxleyi). Our findings suggest that the presence of URA on cellular surfaces of S. costatum (i.e., attached EPS) and its high intracellular fraction could be an indicator for Fe transport from the surrounding seawater to the diatom cells. However, for the coccolithophore E. huxleyi, Fe was not efficiently taken up during its growth. Instead, the more hydrophilic non-attached EPS (i.e., low protein/TCHO ratio) produced by E. huxleyi could have stabilized Fe in the colloidal form as Fe-EPS complexes. Similar partitioning behavior of Fe and Pu suggests that Pu isotopes can potentially serve as a tracer for Fe biogeochemistry in the ocean.

Partitioning of Plutonium by Acid Split Method with Dissolver Solution Derived from Irradiated Fast Reactor Fuel with High Concentration of Plutonium

Partitioning of Plutonium by Acid Split Method with Dissolver Solution Derived from Irradiated Fast Reactor Fuel with High Concentration of Plutonium

Application of a One-Dimensional Transient Electrorefiner Model to Predict Partitioning of Plutonium from Curium in a Pyrochemical Spent Fuel Treatment Process

It has previously been proposed by safeguards experts that curium will track plutonium through a spent fuel pyroprocessing facility, enabling nondestructive assaying of plutonium via counting neutron emissions from 244Cm. This is a critical assumption for the neutron balance approach to safeguards. If Cm and Pu were to behave chemically the same, counting neutrons could be used to estimate Pu concentrations. In this study, plutonium tracking with curium has been investigated using Enhanced REFIN with Anodic Dissolution (ERAD), a one-dimensional transient electrorefiner model based on fundamental electrochemical equations. The model was used to simulate simultaneous deposition of uranium, plutonium, and curium onto a solid metal cathode. Chemical/physical properties used by the model were either obtained from the literature or assumed. The standard exchange current density of curium was estimated by analyzing published cyclic voltammetry data for LiCl-KCl-CmCl3. The focus of the ERAD calculations was on verifying that Pu and Cm could codeposit onto the cathode along with U and to determine if the Pu/Cm ratio would be the same between the salt pool and cathode deposit. It was determined that Cm largely resists cathode deposition, while Pu can be driven to codeposit at sufficiently high current densities. The expected concentration of Cm in the salt would not support any deposition of Cm onto the cathode. It would need to be raised to ~1 wt% before small gram quantities of Cm will deposit onto the cathode. Even then, the Pu/Cm ratio of the cathode was found to be three orders of magnitude higher than the ratio in the salt. It is, thus, concluded that the neutron balance approach would be ineffective at safeguarding a nuclear fuel pyroprocessing facility.

Electrolytic partitioning of uranium and plutonium based on a new type of electrolytic mixer-settler

Summary The design of a new type of electroreduction mixer-settler for the partitioning of uranium and plutonium during the Purex process, which is featured with E-shaped cathodes and U-shaped anodes in settling chamber, is described and the operational results achieved using this equipment are presented. The results show that this new type of mixer-settler has excellent separation performances. The flow rate of organic feed solution is 3 mL/min and the flow ratio of feed solution (1BF) to aqueous back extraction stream (1BX) and to organic wash stream (1BS) is 4/1/1. For an organic feed of 84 g/L uranium and 1.40 ⁓ 2.64 g/L plutonium, both the separation factor of plutonium from uranium and that of uranium from plutonium are apparently higher than 104

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0–8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0–8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5–7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing.

Thermal decomposition studies of aqueous and nitric solutions of hydroxyurea

Hydroxyurea and its derivatives are important nonsalt forming reductants in partitioning of uranium and plutonium in the nuclear fuel reprocessing operations. There is no experimental data available in open literature describing pressurization due to the thermal decomposition of aqueous and nitric solutions of hydroxyurea at elevated temperatures. Authors studied thermal decomposition of hydroxyurea-nitric acid system and resultant pressurization at various concentrations of nitric acid in an adiabatic calorimeter in closed-vent conditions. During these experiments, pressurization was observed. In this paper, results of these experiments have been discussed.

Process modeling of in-situ electrochemical partitioning of uranium and plutonium in purex process: benchmark results with uranium reduction

Abstract In-situ reduction of plutonium and uranium for the separation of U/Pu is suitable for plutonium-rich fuels such as FBR fuels. The mathematical basis for a computer program PUSEP (Plutonium Uranium Solvent Extraction Program) for the analysis of partitioning cycle of PUREX process involving in-situ electrochemical reduction of uranium and plutonium is described in the present investigation. Model equations have been developed on the basis of the idealized model for mixer settlers incorporating distribution coefficients and redox reactions of the species involved and solved numerically to obtain concentration profiles of components. The validity of the model equations and associated computer program is tested by carrying out experiments in a proto type 20-stage electrolytic ejector mixer-settler operating without diaphragm for the electro reduction of uranium. The stage-wise experimental concentration profiles of U(VI), U(IV) and nitric acid were obtained and compared with the theoretical predictions. A reasonably good agreement is achieved between experimental and predicted concentration profiles.

Process modeling of in-situ electrochemical partitioning of uranium and plutonium in purex process: benchmark results with uranium reduction

Process modeling of in-situ electrochemical partitioning of uranium and plutonium in purex process: benchmark results with uranium reduction

Partitioning of plutonium and uranium in aqueous medium using hydroxyurea as reducing agent

A new process for the partitioning of plutonium and uranium during the reprocessing of spent fuel discharged from fast reactor was optimised using hydroxyurea (HU) as a reductant. Stoichiometric ratio of HU required for the reduction of Pu(IV) was studied. The effect of concentration of uranium, plutonium and acidity on the distribution ratio (Kd) of Pu in the presence of HU was studied. The effect of HU in further purification of Pu such as solvent extraction and precipitation of plutonium as oxalate was also studied. The results of the study indicate that Pu and U can be separated from each other using HU as reductant.

Studies on testing and evaluation of platinum plated titanium electrode for production of uranous nitrate

Partitioning of plutonium from uranium is an important step in the reprocessing of spent fuel by PUREX process of solvent extraction using 30% TBP-dodecane. This is achieved by selectively reducing the Pu in solution to least extractable trivalent state by uranous nitrate as the reductant. The latter is conventionally produced by electrolytic reduction of uranyl ion in presence of hydrazine nitrate as uranous nitrate stabilizer using Pt-coated titanium as the anode. The anode plating wears out after period of operation thus affecting the process efficiency and hence the quality control testing of platinum plated electrode becomes important. This article describes the use of Beta backscattering method with strontium-90 radioisotope as non-destructive testing tool for measuring the coating thickness of the sample Ti electrode. The surface characteristics and coating morphology were also examined by scanning electron microscope and the micrographs are presented.

Purification and recovery of plutonium from uranium oxide obtained in the fast reactor fuel reprocessing plant using hydroxyurea as reductant

Spent fuel discharged from Fast Breeder Test Reactor (FBTR) in Kalpakkam is being reprocessed by modified plutonium uranium reduction extraction (PUREX) process using 30% TBP (tributylphosphate) as extractant in the presence of heavy normal paraffin (HNP) as diluent. Partitioning of uranium (U) and plutonium (Pu) is carried out using oxalate precipitation method. Uranium oxide product obtained by this method contains appreciable amount of plutonium which has to be recovered. Recovery of plutonium from this uranium oxide product is carried out by reducing Pu to inextractable Pu(III) using hydroxyurea (HU) and then uranium is extracted into 30% TBP. A small amount of Pu which is extracted in the organic phase is stripped back to aqueous phase by scrubbing with scrubbing agent containing 0.1 M HU in 4 M nitric acid. Similarly U and Pu are co-extracted into 30% TBP and then Pu is removed by scrubbing with 0.1 M HU in 4 M nitric acid. Further decontamination from Pu is obtained in the stripping stages. By this method Pu contamination in the uranium oxide is brought from 7300 ppm to 0.4–3 ppm (wt/wt). This uranium product obtained can be handled on table top.

Using sequential extraction techniques to assess the partitioning of plutonium and neptunium-237 from multiple sources in sediments from the Ob River (Siberia)

Sequential extraction techniques have been developed to assess partitioning of anthropogenic radionuclides ( 240Pu, 239Pu, and 237Np), originating from a variety of sources, as identified by using bulk sediment isotopic composition. Sediments were leached sequentially with a series of six chemical treatments designed to approximate different environmental processes that may occur or to selectively extract trace metals contained in different solid phases of the sediments (i.e., exchangeable, reducible, carbonate, organic, acid leachable, and refractory). Results indicate the majority of Pu and Np is similarly distributed within many of the extracted fractions, with the largest percentage (66–97%) of both elements being observed in sediments treated with buffered citrate dithionite (CDB), which targets easily reduced constituents such as Mn and Fe hydrous oxides. While these results do indicate an association of Pu and Np with redox sensitive elements, the environmental implications are unclear given that the CDB treatment is more extreme than naturally occurring conditions. Minor amounts of Np partition differently from Pu in sediments. The NH 4-acetate treatment, which is designed to liberate trace metals that are loosely adsorbed onto the surfaces of sedimentary materials such as hydrated iron oxides and humic substances, or present at exchangeable sites in clay minerals, mobilized ∼12% of the total Np while Pu levels were below detection. The H 2O 2 treatment, which is designed to liberate trace metals bound to organic matter, mobilized ∼8 and ∼1% of Np and Pu, respectively. These results indicate that a minor portion of the total Np may be affected by environmental conditions that have little or no effect on Pu. Between 7 and 24% of the Pu was observed in treatments designed to liberate Pu and Np that are tightly bound to lithogenic phases or refractory silicates. The 240Pu/ 239Pu observed in accessible and refractory fractions ranged between 0.11 and 0.18. With the exception of one sample, the 240Pu/ 239Pu isotope ratios measured in the refractory fractions are essentially identical having a mean ratio value of 0.123 ± 0.001 (1 σ). The 240Pu/ 239Pu ratios provide isotopic evidence that suggests a portion of non-fallout contamination has a refractory nature. The presence of similarly low 240Pu/ 239Pu ratios in refractory fractions of sediments from the Ob and Irtysh Rivers suggests the existence of a source of refractory Pu which is consistent with refractory “hot particles” derived from surface tests at the Semipalitinsk test site.

Partitioning of uranium and plutonium by acetohydroxamic acid

A process has been developed for separation of plutonium from uranium in the partitioning cycle of reprocessing of fast reactor fuel using a new salt free reagent acetohydroxamic acid (AHA). Acetohydroxamic acid is used as a partitioning agent based on its novel behaviour of reductant/complexant. AHA selectively forms a complex with plutonium than uranium which facilitates stripping of plutonium to the aqueous phase leaving uranium in the organic phase. The effect of plutonium and uranium concentration on feed organics and a number of contacts are investigated in order to maximize the separation factor. It is found from the data that almost 99% of plutonium is stripped into the aqueous phase leaving >93% of uranium in the organic phase itself by just giving three contacts.

A process-based model for the partitioning of soluble, suspended particulate and bed sediment fractions of plutonium and caesium in the eastern Irish Sea

A dynamic model of plutonium behaviour in the marine environment has been developed, representing the oxidation state distribution and partitioning of plutonium between the soluble, colloidal, suspended particulate and seabed sediment fractions. With simple re-parameterisation, this model can also be applied to 137Cs. The model, which is calibrated and validated against field data, has been used to predict concentrations of Pu(α) and 137Cs in both water and seabed sediments from the vicinity of the Sellafield Ltd. reprocessing plant in Cumbria, UK. The model predicts that sediment reworking and transport are the key environmental processes as the Sellafield Pu(α) and 137Cs discharge continues to decline. Inventory calculations generated by the model are consistent with previous estimations. For a hypothetical post-discharge scenario, the concentrations of these radionuclides in both seawater and surface sediments are predicted to decrease sharply, concurrent with a downward vertical migration of the activity retained in sediments.

AMERICIUM IN THE BEAGLE DOG: BIOKINETIC AND DOSIMETRIC MODEL

A biokinetic model of the systemic distribution of americium in the beagle dog is presented. The model is based on a previous biokinetic model of plutonium. The data sets used for the development of the model were the measurements of excreted activity (urine and feces) and organ burdens (skeleton, liver, and other soft tissues) for different levels of initial injected activity. In developing the model, the compartmental structure of the skeleton of the plutonium model was adopted, and only the numerical values of parameters were adapted. The model well describes the fractions of americium in the skeleton, liver, and soft tissues and the total fraction excreted in urine and feces. The tuning of the liver clearance parameter provides a realistic description of the change in the partitioning between liver and skeleton for different injection levels. The most significant features of the biokinetics and dosimetry of americium and plutonium in beagles are compared. The total fractions of the clearance to the skeleton and the liver are roughly equal to the value for plutonium, but the partitioning of americium between these organs is reversed with respect to the partitioning of plutonium. 241Am doses to liver and skeleton are similar to 239Pu doses, owing to some counteracting factors. For the highest injection level, the liver mass is dependent on the time post injection. For the skeletal tissues, the dose to the cortical endosteum by far exceeds the dose to the trabecular endosteum and the red marrow. The model provides the basis for statistical survival analyses and risk estimates.

Solid–solution partitioning of plutonium in surface waters at the Atomic Weapons Establishment Aldermaston (UK)

The Atomic Weapons Establishment (AWE) at Aldermaston (Berkshire, UK) has provided and maintained the warheads for the UK's nuclear deterrent for more than 50 years. Whilst the site is radiologically safe, in a few locations the soil contains specific activities of plutonium (Pu) above background arising from a legacy of historic operations. Run-off water (a mixture of rainwater and groundwater) from part of the site is routed into a water management system, and after analysis and radiological assessment, released into local streams. Water and sediment samples have been collected from a number of closely spaced locations within this system to assess the solid–solution partitioning of Pu. Survey work was complemented by batch type desorption experiments to assess redissolution from ‘contaminated’ sediment into ‘uncontaminated’ water. The survey data indicate that specific activities of both dissolved and particle bound 239+240Pu varied by roughly two orders of magnitude, ranging from ∼0.7 μBq kg −1 up to ∼44 μBq kg −1, and ∼1.2 Bq kg −1 up to ∼400 Bq kg −1, respectively, consistent with water originating from different parts of the site. Apparent K d values varied by an order of magnitude (from 0.7–16×10 6) with an average value of 4×10 6. Results from the desorption experiments indicated the extent of redissolution was very small and the derived K d's corroborated values obtained from the survey work. K d's given here are compared with other literature values, and are the greatest reported to date. Results are also provided describing the variation in water quality parameters in shallow groundwater samples. Alkalinity values ranged from 120 to 388 mg l −1 CaCO 3 with an average value of 195 mg l −1 CaCO 3. Corresponding values for pH were 6.6–8.3 with an average of 7.5. Over half of the samples were estimated to be supersaturated with respect to calcite. It is suggested that the state of calcite saturation may be responsible for the high Pu K d. As a consequence of the high particle reactivity, migration of Pu contamination, both within and away from the AWE Aldermaston site, is likely to be very restricted.

Source-term characterisation and solid speciation of plutonium at the Semipalatinsk NTS, Kazakhstan

New data on the concentrations of key fission/activation products and transuranium nuclides in samples of soil and water from the Semipalatinsk Nuclear Test Site are presented and interpreted. Sampling was carried out at Ground Zero, Lake Balapan, the Tel’kem craters and reference locations within the test site boundary well removed from localised sources. Radionuclide ratios have been used to characterise the source term(s) at each of these sites. The geochemical partitioning of plutonium has also been examined and it is shown that the bulk of the plutonium contamination at most of the sites examined is in a highly refractory, non-labile form.

Evaluation of a conceptual model for the subsurface transport of plutonium involving surface mediated reduction of Pu(V) to Pu(IV)

A conceptual model is proposed to explain the transport behavior of plutonium in laboratory columns packed with a sandy coastal soil from the U.S. Department of Energy (DOE)'s Savannah River Site. The column transport experiments involved the introduction of a finite step input of plutonium, predominately in the +5 oxidation state, into the columns followed by elution with a low-carbonate solution of 0.02 M NaClO 4 at pH 3, 5, and 8. Total plutonium concentrations were measured in the effluent as a function of time. These elution profiles suggest at least two distinct physical/chemical forms of plutonium, each with a different mobility. To explain the observed behavior, the following conceptual model was evaluated: [1] equilibrium partitioning of plutonium (V) and plutonium (IV) between the aqueous and sorbed phases as defined by pH-dependent, oxidation-state specific distribution coefficients and [2] kinetic reduction of plutonium (V) to plutonium (IV) in the sorbed phase. The conceptual model was applied to the column experiments through a one-dimensional advective/dispersive mathematical model, and predictions of the mathematical model were compared with the experimental data. Overall, the model was successful in predicting some of the major features observed in the experiments. It also yielded quantitative estimates of the rate constant for surface mediated reduction of plutonium (V) to plutonium (IV) that were of the same order (10 −4 to 10 −5 s −1) as those calculated from batch data both for this soil and for goethite.

Transport of plutonium in surface and sub-surface waters from the Arctic shelf to the North Pole via the Lomonosov Ridge

New data on the levels and long-range transport of plutonium in the Arctic Ocean, recorded in the course of two expeditions to this zone in 1994 and 1996, are discussed in this paper. Specifically, approximately 100 plutonium measurements in surface and sub-surface water sampled at 58 separate stations throughout the Kara, Laptev and East Siberian Seas, as well as along latitudinal transects across the Lomonosov Ridge, are reported and interpreted in terms of the circulation pathways responsible for the transport of this element from the North Atlantic to the Arctic Shelf and into the Arctic interior. In addition, the behaviour of plutonium in its transit through the vast Arctic shelf seas to open waters under extreme environmental conditions is discussed in terms of the partitioning of plutonium between filtered (<0.45 μm) seawater and suspended particulate, and its association with colloidal matter. Finally, limited evidence of the presence of a colloidal plutonium component in Arctic waters subject to direct riverine input is adduced.

Partitioning of plutonium between glass, secondary phases, and contact refractories

AbstractA wide variety of plutonium-containing wastes have been generated from both defense and commercial nuclear programs. The Advanced Mixed Waste Treatment Project (AMWTP) involves the privatized treatment of 65,000 m3 of mixed TRU and low-level waste that is in storage at DOE's INEEL site. BNFL Inc. leads the commercial team including GTS Duratek, Inc., that has been selected to design, construct, and operate a facility to perform that service. Since vitrification is a potential treatment option, the current study was conducted to investigate issues surrounding the incorporation of the approximately 700 kg of plutonium that is present in the AMWTP waste. A series of glasses were prepared with plutonium (0.5 wt%), with cerium (as a potential surrogate), with both, and with neither. Scanning electron microscopy coupled with energy dispersive x-ray spectroscopy (SEM-EDS) and optical analyses confirmed that both elements were completely dissolved into the as-melted glasses. Heat treatments forced the formation of secondary phases, which were analyzed for Pu and Ce incorporation. Plutonium-containing glasses were spiked with sulfates and remelted to force the formation of a sulfate salt phase, which was analyzed to determine partitioning of Pu and Ce. Coupons of Monofrax K-3 glass contact refractory were exposed to melts of these glasses for six days at 1208°C using both standard and modified versions of ASTM procedure C621-84. Thickness loss, microstructural alteration, and the extent to which Pu and Ce were incorporated into the material were then determined by methods employing SEMEDS\ wavelength dispersive x-ray spectroscopy (WDS). Extreme heat treatments were found to lead to the formation of small amounts (&lt; 0.1 vol%) of either (Pu, Ce)O2, or PuO2, depending upon whether Ce was present or absent, respectively. The Pu level in the sulfate phase was at least a factor often below that of the parent glass, and no tendency to concentrate Pu in the K-3 refractory above the level in the parent glass was found.