Uranium Plutonium Research Articles

Characterization of self-damaged (U,Pu)N fuel used in the NIMPHE program

During in-pile, irradiation-induced damage occurs in nuclear fuel and results in a deterioration of its properties, which can affect the margin to melt of the fuel. Damage also occurs in fresh fuel through the self-irradiation process, and thus provides a convenient means to investigate changes in the material. A uranium–plutonium mixed nitride fuel made over 25years ago, and stored in the ITU archives has been retrieved. Coupling EXAFS and TEM has shown that this material was still well-crystallized. However, an increase of 0.3% of the lattice parameter was found. As shown by the EXAFS, the U–N and Pu–N as well as the metal–metal distances are similarly affected. However, no significant modification of both anion and cation sublattice was found. Although no defect clustering was found by EXAFS, the presence of nanometric helium bubbles was demonstrated by TEM as well as nanometric disordered domains.

Uncertainty quantification of neutronic parameters of light water reactor fuel cells with JENDL-4.0 covariance data

Neutronic parameter uncertainty induced by nuclear data uncertainty is quantified for several light water reactor fuel cells composed of different combinations of fissile/fertile nuclides. The covariance data given in JENDL-4.0 are used as the nuclear data uncertainty, and uncertainty propagation calculations are carried out using sensitivity coefficients calculated with the generalized perturbation theory for burnup-related neutronic parameters. It is found that main contributors of nuclear data uncertainty to the neutronic parameter uncertainty are the uranium-238 capture cross section in a uranium-oxide fuel cell, and the plutonium-240 and plutonium-241 capture cross sections and fission spectrum of fissile plutonium isotopes in a uranium–plutonium mixed-oxide fuel cell. It is also found that thorium-232 capture cross section uncertainty is a dominant source of neutronic parameter uncertainty in thorium–uranium and thorium–plutonium mixed-oxide fuel cells. It should be emphasized that precise and detail information of component-wise uncertainties can be obtained by virtue of the adjoint-based sensitivity calculation methodology. Furthermore, cross-correlations are evaluated for each fuel cell, and strong correlations among the same parameters at the beginning of cycle and at the end of cycle and among different parameters are observed.

Progress towards the Full Recovery of Neptunium in an Advanced PUREX Process

To meet the needs of future closed fuel cycles, the complete recovery of minor actinides, including neptunium, may be required. Neptunium can be fully recovered by modifications to the Plutonium URanium Extraction (PUREX) process but this requires careful control of the Np(V)-(VI) redox reaction in the first solvent extraction contactor to avoid losses to the highly active aqueous raffinate, as occurs in current reprocessing plants. As part of the on-going development of an Advanced PUREX process we report a series of solvent extraction experiments aimed at optimizing neptunium recovery in a process that is based on centrifugal contactors as the extraction equipment. Suitable experimental conditions for Np(V) oxidation were identified through simple stirred 2-phase experiments and single stage mini-centrifugal contactor experiments. A U/Np-active proof-of-principle flowsheet test in a multi-stage centrifugal contactor cascade then demonstrated > 99% extraction of neptunium, thus suggesting the aims for neptunium recovery in advanced fuel cycles can be met by an Advanced PUREX process.

Curium analysis in plutonium uranium mixed oxide by x-ray fluorescence and absorption fine structure spectroscopy

Plutonium uranium mixed oxide (MOX) fuels are being used in commercial nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regards to their environment and the coolant. In this work the study of the local occurrence, speciation and next-neighbour environment of curium (Cm) in the (Pu,U)O2 lattice within an irradiated (60MWdkg−1 average burn-up) MOX sample was performed employing micro-x-ray fluorescence (µ-XRF) and micro-x-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Cm (∼0.7wt% in the rim and ∼0.03wt% in the centre) are determined from the experimental data gained for the irradiated fuel material examined in its centre and peripheral (rim) zones of the fuel. Curium occurrence is also reduced from the centre (hot) to the periphery (colder) because of the condensation of these volatile oxides. In the irradiated sample Cm builds up as Cm3+ species (>90%) within a [CmO8]13− or [CmO7]11− coordination environment and no (<10%) Cm(IV) can be detected in the rim zone. Curium dioxide is reduced because of the redox buffering activity of the uranium dioxide matrix and of its thermodynamic instability.

Reduction of uranium (VI to IV) by hydrogenation using Adams’ catalyst

Plutonium uranium reduction extraction using U(IV) as universal reductant for Pu partitioning is the only technology practiced internationally to recover U and Pu from spent nuclear fuels. Uranous requirement of Indian reprocessing plants is met by the electrolytic reduction of uranyl nitrate with 50–60% conversion. Though the current requirement can be met with this method, it increases the load on uranium purification cycle. In addition, it is a batch process with slow kinetics. In order to achieve higher conversion of uranyl nitrate to uranous nitrate, catalytic reduction method using hydrogen in presence of Adams’ catalyst (PtO2) was tried. Parametric studies have been performed in an autoclave to evaluate the effect of U(VI) concentration, the role of hydrazine nitrate and pressure. It is observed that kinetics is improved at higher pressures. The studies revealed that near total conversion of uranium from (VI to IV) can be achieved by the catalytic reduction route.

Determination of the oxygen potential of U0.718Ce0.282O2 ± x using a solid electrolyte galvanic cell

To assess the nature of variations in the thermodynamic properties of uranium plutonium oxide fuel, which is highly radioactive and toxic, we have measured characteristics of uranium cerium oxides similar in properties to the nuclear fuel. The emf method, a solid electrolyte galvanic cell, and solid-state coulometric titration have been used to find out general trends in the variation of the oxygen potential with the oxygen/metal ratio for U0.718Ce0.282O2 ± x .

Sustainability issues of plutonium recycling in light water reactors: Code evaluations up to 2050

Plutonium recycling in light water reactors is a viable and mature technology capable of improving several indicators of great importance in the evaluation of the sustainability of nuclear energy development in the near- and long-term such as the shortage of natural uranium resources, the increase of spent fuel inventories and the proliferation risks of high-level radioactive waste. This paper, after a brief review of the status of plutonium recycling, presents the results of a scenario analysis carried out in the hypothesis that the share of nuclear fleet loaded with mixed uranium–plutonium oxide fuel (MOX) is kept constant up to the middle of the century. Beside mentioned indicators, the paper discusses the needs for complex and costly fuel cycle infrastructures required for the reprocessing of spent nuclear fuel and the fabrication of MOX. Assuming that the deployment of fast reactors occurs beyond the middle of the century, the article focuses on the comparison of an open fuel cycle strategy with a closed fuel cycle strategy where plutonium is recycled.Presented calculations, according to moderate and high projections of nuclear energy development, confirm that plutonium recycling, although deployed to a limited extent, could be beneficial in reducing the stockpiles of nuclear spent fuel and in reducing the risks of proliferation due to the amount of fissile plutonium in the system. The improvement found in the consumption of natural uranium resources was limited promoting to this purpose the deployment of next-generation fast reactors. If a moderate development of nuclear energy is confirmed, the current capacity for reprocessing and MOX fuel fabrication could be sufficient to cope with the foreseen demand, on the contrary, significant investments could be necessary in case of steep increase of installed nuclear energy. Calculations were performed by means of the DESAE code (Dynamic Energy System – Atomic Energy), a tool developed within the IAEA INPRO project.

Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O₂ lattice in an irradiated (60 MW d kg⁻¹) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (~0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am³⁺ species within an [AmO₈]¹³⁻ coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix. - Graphical abstract: Americium LIII XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 μm×300 μm beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Ammore » is found as trivalent in the UO₂ matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: • Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. • The americium redox state as determined from XAS data of irradiated fuel material was Am(III). • In the sample, the Am³⁺ face an AmO₈¹³⁻coordination environment in the (Pu,U)O₂ matrix. • The americium dioxide is reduced by the uranium dioxide matrix.« less

Direct Selective Extraction of Actinides (III) from PUREX Raffinate using a Mixture of CyMe4BTBP and TODGA as 1-cycle SANEX Solvent Part III: Demonstration of a Laboratory-Scale Counter-Current Centrifugal Contactor Process

The direct selective separation of the trivalent actinides americium and curium from a simulated Plutonium Uranium Refining by EXtraction (PUREX) raffinate solution by a continuous counter-current solvent extraction process using miniature annular centrifugal contactors was demonstrated on a laboratory scale. In a 32-stage spiked test (12 stages for extraction, 16 stages for scrubbing, and 4 stages for Am/Cm stripping), an extractant mixture of CyMe4BTBP and TODGA in a TPH/1-octanol mixture was used. The co-extraction of some fission and corrosion product elements, such as zirconium and molybdenum, was prevented by using oxalic acid. Co-extracted palladium was selectively stripped using an L-cysteine scrubbing solution and the trivalent actinides were selectively stripped using a glycolic acid-based stripping solution. It was demonstrated that a selective extraction and high recovery of > 99.4% of the trivalent minor actinides was achieved with low contamination by fission and corrosion products. The product contained 99.8% of the initial americium and 99.4% of the initial curium content. The spent solvent still contained high concentrations of Cu, Cd, and Ni. The experimental steady-state concentration profiles of important solutes were determined and compared with those from computer-code calculations.

Simple formula to evaluate helium production amount in fast reactor MA-containing MOX fuel and its accuracy

A simple formula is developed for the evaluation of the helium production amount in fast reactor minor actinide (MA) containing uranium–plutonium mixed oxide (MOX) fuel. For the subroutine use in the existing fuel behavior analysis code, the formula is designed putting emphasis on simplicity and quickness rather than accuracy. The accuracy of the formula is confirmed by comparing with the detailed calculation with SWAT code, and also with the post irradiation examination (PIE) results of the fuel pin irradiated at the experimental fast reactor JOYO. As a result, it is found that the formula evaluates the helium production amount with the difference of less than about 10% from the detailed calculation and the PIE results, when the MA isotope content is less than 5 wt.%. Based on these results, the formula is installed in the fuel behavior analysis code for the simulation of helium behavior in fast reactor fuels.

Direct comparison of spectroscopic data with cluster calculations of plutonium dioxide and uranium dioxide

The density of states predicted by cluster calculations of PuO2 and UO2 is directly compared to the results from soft x-ray spectroscopy. Remarkably good agreement between the experimentally measured spectroscopic peaks and the calculated density of states of the central part of the cluster is observed.

Characterization of polymeric films subjected to lithium ion beam irradiation

Two different polymeric materials that are candidate materials for use as binders for mixed uranium–plutonium oxide nuclear fuel pellets were subjected to Li ion beam irradiation, in order to simulate intense alpha irradiation. The materials (a polyethylene glycol 8000 and a microcrystalline wax) were then analyzed using a combination of mass spectrometry (MS) approaches and X-ray photoelectron spectroscopy (XPS). Samples of the irradiated PEG materials were dissolved in H2O and then analyzed using electrospray ionization-MS, which showed the formation of a series of small oligomers in addition to intact large PEG oligomers. The small oligomers were likely formed by radiation-induced homolytic scissions of the C–O and C–C bonds, which furnish radical intermediates that react by radical recombination with H and OH. Surface analysis using SIMS revealed a heterogeneous surface that contained not only PEG-derived polymers, but also hydrocarbon-based entities that are likely surface contaminants. XPS of the irradiated PEG samples indicated the emergence of different carbon species, with peak shifts suggesting the presence of sp2 carbon atoms. Analysis of the paraffinic film using XPS showed the emergence of oxygen on the surface of the sample, and also a broadening and shifting of the C1s peak, demonstrating a change in the chemistry on the surface. The paraffinic film did not dissolve in either H2O or a H2O–methanol solution, and hence the bulk of the material could not be analyzed using electrospray. However a series of oligomers was leached from the bulk material that produced ion series in the ESI-MS analyses that were identified octylphenyl ethoxylate oligomers. Upon Li ion bombardment, these shifted to a lower average molecular weight, but more importantly showed the emergence of three new ion series that are being formed as a result of radiation damage. Surface analysis of the paraffinic polymers using SIMS produced spectra that were wholly dominated by hydrocarbon ion series, and no difference was observed between unirradiated and irradiated samples. The studies demonstrate that for the PEG-based polymers, direct evidence for radiolytic scission can be observed using ESI-MS, and suggests that both radiolytic pathways and efficiencies as a function of dose should be measurable by calibrating instrument response to the small oligomeric degradation products.

Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71.

A physical model for evaluating uranium nitride specific heat

Nitride fuel is one of perspective materials for the nuclear industry. But unlike the oxide and carbide uranium and mixed uranium–plutonium fuel, the nitride fuel is less studied. The present article is devoted to the development of a model for calculating UN specific heat on the basis of phonon spectrum data within the solid state theory.

Extraction Studies of Phosphoric Acid Tributyl Ester from Nitric Acid Solutions Using Dodecane in the Presence of Inextractable Metal Nitrates

Sodium nitrate and calcium nitrate are not extracted by phosphoric acid tributyl ester (TBP) in the plutonium uranium extraction (PUREX) process. The effect of these inextractable metal nitrates on the solubility and distribution data of TBP has been studied. The solubility of TBP in water in the presence of different concentrations of sodium nitrate is determined, and salting coefficient of TBP is calculated. It has been observed that the solubility of TBP in nitric acid decreases in the presence of both inextractable nitrates, that is, sodium nitrate and calcium nitrate. The effect of inextractable metal nitrates on the equilibrium data for the TBP–dodecane–nitric acid system has been studied. The distribution coefficient value of TBP is greater in the presence of sodium nitrate solutions as compared with calcium nitrate solutions due to more salting out of TBP. Because sodium nitrate is a better salting agent than calcium nitrate, the variation in the concentration of nitric acid at equilibrium was more in sodium nitrate solutions in comparison with that in calcium nitrate solutions. The data obtained will be useful in separating dissolved TBP from acidic waste for safe disposal of nuclear waste.

Modeling of Pu(VI) distribution coefficients in 30 % TBP based PUREX process

Hexavalent plutonium (Pu(VI)) is an important solute in the PUREX (plutonium uranium extraction) process. In 30 % TBP based PUREX solvent extraction system, distribution coefficient of Pu(VI) is much lower than that of Pu(IV). This lower distribution coefficient of Pu(VI) may cause unexpected Pu loss during primary HA extraction in low acid flowsheets. An empirical model for Pu(VI) distribution coefficients in 30 % TBP and its temperature dependency has been reported in this paper. Comparison with literature data revealed a reasonably good agreement between the reported experimental and model predicted values.

Studies on solubility of TBP in aqueous solutions of fuel reprocessing

In nuclear technology, tri-n-butyl phosphate (TBP) diluted with a hydrocarbon diluent such as n-dodecane or NPH is the most frequently used solvent in liquid–liquid extraction for fuel reprocessing. This extraction, known as the plutonium uranium refining by extraction, is still considered as the most dominant process for the extraction of uranium and plutonium from irradiated fuels. The solubility of pure TBP in water is about 0.4 g/L at 25 °C. This is enough to create trouble during evaporation of raffinate and product solutions. Solubility data for undiluted TBP and TBP (diluted in inert hydrocarbon diluent) in various concentrations of nitric acid is not adequate in the literature. The solubility data generated in the present study provide complete information on the solubility of TBP in various nitric acid concentrations (0–15.7 M) at room temperature. The effect of heavy metal ion concentration such as uranium and various fission products on the solubility of TBP in nitric acid is also presented. The results obtained from gas chromatographic technique were compared with spectrophotometric technique by converting the organic phosphate into inorganic phosphate. The generated data is of direct relevance to reprocessing applications.

Study of thorium–uranium based molten salt blanket in a fusion–fission hybrid reactor

Not only solid fuels, but also liquid fuels can be used for the fusion–fission symbiotic reactor blanket. The operational record of the molten salt reactor with F–Li–Be was very successful, so the F–Li–Be blanket was chosen for research. The molten salt has several features which are suited for the fusion–fission applications.The fuel material uranium and thorium were dissolved in the F–Li–Be molten salt. A combined program, COUPLE, was used for neutronics analysis of the molten salt blanket. Several cases have been calculated and compared. Not only the influence of the different fuels have been studied, but also the thickness of the molten salt, and the concentration of the 6Li in the molten salt.Preliminary studies indicate that when thorium–uranium–plutonium fuels were added into a F–Li–Be molten salt blanket and with a component of 71% LiF–2% BeF2–13.5% ThF4–8.5% UF4–5% PuF3, and also with the molten salt thickness of 40cm and natural concentration of 6Li, the appropriate blanket energy multiplication factor and TBR can be obtained.The result shows that thorium–uranium molten salt can be used in the blanket of a fusion–fission symbiotic reactor. The research on the molten salt blanket must be valuable for the design of fusion–fission symbiotic reactor.

Une curieuse carcinose péritonéale

Nitrides have been proposed to be a suitable material for fast neutronic systems from beginning of the development of nuclear fuel. Starting with the production of uranium nitride and sesquinitrides up to mixed plutonium uranium nitrides, today's developments are inert nitride matrix materials to burn plutonium or to transmute long-lived actinides in accelerator-driven sub-critical systems (ADS) or fast reactors (FR). Several authors proposed zirconium nitride as possible inert matrix material for this reason. Mixed zirconium nitrides can be fabricated by carbothermal nitridation of the oxides in a narrow temperature window. Obtaining high quality material with low carbon and oxygen content is still the major challenge. Producing mixed nitride fuels by special shaping methods, as for example direct coagulation casting or freeze drying, in comparison to conventional powder compaction enables to use this material in special shapes to optimize burn up.The history of nitride fuels at PSI will be shown up to today's CONFIRM project, dealing with plutonium zirconium nitride fuels.

Fundamental solution of nuclear solitary wave

This paper deals with the problem of asymptotic breeding/burning waves during long term nuclear fission processes. The uranium–plutonium (U–Pu) conversion cycle is considered under fast spectrum conditions. A one-group diffusion equation coupled with burn-up equations is set up. The nuclide atom number densities can be determined as functions of the neutron fluence only, as the natural radioactive processes are neglected. It is found then that the diffusion equation with the neutron fluence dependent macroscopic cross sections is analytically integrable (solvable) in the 1-D case without feedback effects. A permanent solitary wave solution exists under certain conditions, where the infinite medium multiplication factor first increases from a subcritical level up to a supercritical point and then falls to another subcritical level again, along with the increasing neutron fluence. Relationships between wave amplitude, wave number, parameters of fuel are studied. A representative example is shown for a breeding/burning solitary wave propagating in a 238U medium with a suitable content of burnable poison, where the conversion chain is considered up to 242Pu. Finally it is demonstrated as well that in a two-dimensional cylindrical case, a multi-dimensional permanent solitary wave pattern with a constant drift speed can be also achieved based on the 1-D fundamental solution by adjusting the initial radial distribution of the fuel composition, where more fuel and higher enrichment are needed in the outer peripheral region than in the inner one.