Applications In Nuclear Fuel Cycle Research Articles

Pu(VI) Oxalate Crystal Structure and Evidence of Photoreduction to Pu(IV) Oxalate.

We report the first crystal structure of a Pu(VI)-oxalate compound. This compound, [PuO2(C2O4)(H2O)]·2(H2O) (1), crystallizes in space group P21/c with a = 5.5993(3) Å, b = 16.8797(12) Å, c = 9.3886(6) Å, and β = 98.713(6)°. It is isostructural with the previously reported U(VI) compound, [UO2(C2O4)(H2O)]·2(H2O). Each plutonyl ion (PuO22+) is coordinated in the equatorial plane by two side-on bidentate oxalates, creating an infinite chain along [001]. A coordinated water molecule and twisting of the oxalates lead to a distorted pentagonal bipyramidal geometry of the Pu. A photochemical degradation was observed for 1, which resulted in the formation of a secondary crystalline phase. The absorption spectrum of this secondary phase confirmed the presence of Pu(IV), but it did not match the spectrum of Pu(C2O4)2·6H2O, which is considered to be the primary product of Pu-oxalate precipitation. While compound 1 has previously been proposed to exist in solution, this is the first time it has been isolated via crystallization. Although redox interactions between Pu and oxalate have been documented in the literature, the present study is the first observation of a photochemical reduction of Pu(VI)-oxalate. As a result, this study has expanded on the limited understanding of the Pu(VI)-oxalate system, which is important for nuclear fuel cycle applications.

Radiation stability testing of hydrated and heat-treated magnesium potassium phosphates for back end nuclear fuel cycle applications

Magnesium potassium-based phosphate (MKP) cement MgKPO4.6H2O is proposed as a promising alternative for immobilizing nuclear waste. In order to assess the long-term performance of MKP based cementitious matrices under harsh service conditions, it is important to investigate the radiation effects on the material at doses representative of service environment. In the present study, MKP cement samples were prepared by gel crystallization method and subsequently heat treated at 120 °C,400 °C,800 °C and 1000 °C. The powders of as-prepared and annealed MKP specimens were subjected to 10 MeV electron beam irradiation for 20 MGy dose. Electron irradiation resulted in the partial removal of structural water from the structure of as-prepared MKP specimen as confirmed through vibrational spectroscopy techniques. Amongst the as-prepared and annealed MKP specimens, MKP-1000 showed best radiation stability. Detailed insights on the radiation stability of these matrices are discussed.

In View of “On-Site” Nuclear Forensics and Assay of Fissile Materials in Sealed Packages by High-Resolution γ-Ray Spectrometry

Several incidences of nuclear smuggling during the past few decades have raised the demand for the development of a strong "on-site" nuclear forensic infrastructure. High-resolution γ-ray spectrometry (HRGRS) plays an important role in nuclear forensics. However, the existing methodologies, developed primarily for nuclear fuel cycle applications, are relative and rely on the availability of a standard, limiting their use for the absolute assay of special nuclear materials in nonstandard geometry samples with an unknown matrix, which is vital to make a quick "on-site" decision on the severity, potential radiological threat, and intended use of an interdicted package. In this work, a methodology has been developed using HRGRS for quantifying fissile (235U, 239Pu) and other radioisotopes, which is applicable to sealed packages without requiring the knowledge of the sample geometry and the matrices. By combining experiments and Monte Carlo simulations, an iterative methodology has been proposed for "point" to "extended" source absolute efficiency transformation and demonstrated further for the absolute isotopic assay of uranium and plutonium standards, mock-up nuclear forensic samples, and an unknown nuclear material mixture with a nonstandard geometry, compound matrices, and a wide variation in the elemental and isotopic compositions with a view to imitate an "on-site" experience. The present methodology requires an assay time of only a few minutes to an hour and thus promises "on-site" nuclear forensic analysis of suspected flagged packages at borders and ports using high-resolution γ-ray spectrometry. Furthermore, the present methodology is versatile and can also be adopted for wider applications, beyond nuclear forensics.

Simultaneous quantification of uranium(VI), samarium, nitric acid, and temperature with combined ensemble learning, laser fluorescence, and Raman scattering for real-time monitoring.

Laser-induced fluorescence spectroscopy (LIFS), Raman spectroscopy, and a stacked regression ensemble was developed for near real-time quantification of uranium(VI) (1-100 μg mL-1), samarium (0-200 μg mL-1) and nitric acid (0.1-4 M) with varying temperature (20 °C-45 °C). LIFS applications range from fundamental lab-scale studies to real-time process monitoring at industrial levels, such as nuclear reprocessing applications, provided the phenomena affecting the fluorescence spectrum are accounted for (e.g., absorption, quenching, complexation). Multiple chemometric models were examined and compared to a more traditional multivariate regression approach called partial least squares (PLS). Results obtained on synthetic samples selected using D-optimal experimental design indicated that a stacked regression method, which included ridge regression, random forest, PLS, and an eXtreme gradient boost algorithm, successfully measured uranium(VI) concentrations directly in nitric acid without measuring luminescence lifetimes or standard addition. The top model resulted in percent root-mean-square error of prediction values of 5.2, 1.9, 3.0, and 2.3% for U(VI), Sm3+, HNO3, and temperature, respectively. The approach may be useful for quantifying fluorescent fission products (e.g., Sm3+) to provide information on burnup of irradiated nuclear fuel. This novel framework reinforces the applicability of LIFS for real-time applications in nuclear fuel cycle applications.

Highly efficient and selective extraction of Pu(IV) using two alkyl-substituted amides of nitrilo triacetic acid from nitric acid solutions

• The extraction of U(VI), Pu(IV) and Am(III) was studied with two tripodal NTA amide ligands. • The extraction trend was Pu(IV) > >U(VI) > Am(III) with both the ligands and HONTA > HDDNTA. • The metal ion extraction was not significantly affected with U loading or absorbed gamma dose. • Stripping was facile and the encouraging reusability data promises for possible application. • Modifier content has a positive impact on water and nitric acid extraction and negative impact on metal ion extraction. Liquid-liquid extraction of the actinide ions UO 2 2+ , Pu 4+ , Am 3+ was carried out from nitric acid feed solutions using two tripodal amide extractants, viz., N,N,N’,N’,N”,N” - hexa - n -octylnitrilotriacetamide (HONTA) and N,N,N’,N’,N”,N” - hexa - n -dodecylnitrilotriacetamide (HDDNTA) in 10% isodecanol + 90% n -dodecane diluent. The metal ion extraction trend was: Pu(IV) ≫ U(VI) > Am(III) at 3 M HNO 3 , HONTA being a superior extractant than HDDNTA. Slope analysis indicated ML 2 type species for both Pu(IV) and Am(III) at 3 M HNO 3 and pH 2.0, respectively, whereas, only in case of HONTA, a ML type species was obtained with U(VI) at 3 M HNO 3 . Back extraction data suggested that a mixture of 0.5 M HNO 3 + 0.5 M oxalic acid was efficient for quantitative stripping (∼90%) of Pu(IV), whereas 1 M Na 2 CO 3 and 1 M α-HIBA gave satisfactory back extraction for U(VI) and Am(III), respectively. The ligand solutions showed reasonably good radiation stability up to an absorbed total gamma ray dose of 274 kGy. Distribution data were also obtained under U(VI) loading conditions for possible nuclear fuel cycle applications.

Hollow fibre supported liquid membranes for nuclear fuel cycle applications: A review

This article gives a comprehensive review on hollow fibre supported liquid membrane based methods for the separation of actinides, lanthanides and fission elements in nuclear fuel cycle applications. The nuclear fuel cycle operations result in large volumes wastes containing long-lived actinides and fission elements which need to be remediated before their release into the environment. The routinely used methods for such wastes treatments are precipitation, solvent extraction and ion-exchange. However, these traditional methods lead to solid waste generation, use of large volumes of toxic/inflammable chemicals and suffer with slow processing issues. In the past few decades, attempts have been made to explore the supported liquid membrane based separation methods using hollow fibre contactors with highly encouraging results. This article gives a comprehensive account of studies on hollow fibre supported liquid membrane aimed at separations relevant to the back-end of nuclear fuel cycle.

Actinide ion extraction using room temperature ionic liquids: opportunities and challenges for nuclear fuel cycle applications.

Studies on the extraction of actinide ions from radioactive feeds have great relevance in nuclear fuel cycle activities, mainly in the back end processes focused on reprocessing and waste management. Room temperature ionic liquid (RTIL) based diluents are becoming increasingly popular due to factors such as more efficient extraction vis-à-vis molecular diluents, higher metal loading, higher radiation resistance, etc. The fascinating chemistry of the actinide ions in RTIL based solvent systems due to complex extraction mechanisms makes it a challenging area of research. By the suitable tuning of the cationic and anionic parts of the ionic liquids, their physical properties such as density, dielectric constant and viscosity can be changed which are considered key parameters in metal ion extraction. Aqueous solubility of the RTILs, which can lead to significant loss in the solvent inventory, can be avoided by appending the extractant moieties onto the ionic liquid. While the low vapour pressure and non-flammability of the ionic liquids make them appear as 'green' diluents, their aqueous solubility raises concerns of environmental hazards. The present article gives a summary of studies carried out on actinide ion extraction and presents perspectives of its applications in the nuclear fuel cycle. The article discusses various extractants used for actinide ion extraction and at many places, comparison is made vis-à-vis molecular diluents which includes the nature of the extracted species and the mechanism of extraction. Results of studies on rare earth elements are also included in view of their similarities with the trivalent minor actinides.

Acid-switched Eu(III) coordination inside reverse aggregates: Insights into a synergistic liquid-liquid extraction system

Determining the structure of complex solutions bearing metal ions is challenging, but crucial for developing important technologies such as liquid-liquid extraction for metal refining and separation purposes. Herein, the structure of an organic Eu(III) solution consisting a binary mixture of lipophilic ligands di-2-ethylhexyl phosphoric acid (HDEHP) and tetraoctyl diglycolamide (TODGA) in dodecane is studied using synchrotron small angle X-ray scattering (SAXS) and X-ray absorption fine structure spectroscopy (EXAFS). This system is of technological importance in f-element separation for nuclear fuel cycle applications, where extraction is controlled by varying nitric acid concentration. Extraction is promoted at low and high concentration, but is retarded at intermediate concentration, leading to a U-shaped function; the structural origins of which we investigate. At the nanoscale, the solution is apparently comprised of reverse micelles with polar cores of approximately 1nm in size, and these remain virtually unchanged as acid concentration is varied. Inside the polar cores, the coordination environment of Eu(III) switches from a 9-coordinate [Eu(TODGA)3]3+ motif at high acid, to a 6-coordinate HDEHP-dominated complex resembling Eu(HDEHP·DEHP)3 at low acid. The results show that extraction is controlled within the coordination sphere, where it is promoted under conditions that favor coordination of either one of the two organic ligands, but is retarded under conditions that encourage mixed complexes. Our results link solution structure with ion transport properties in a technologically-important liquid-liquid ion extraction system.

Accelerator–Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of accelerator-driven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

Characterizing Diamylamylphosphonate (DAAP) as an Americium Ligand for Nuclear Fuel-Cycle Applications

Successful deployment of the currently-envisioned advanced nuclear fuel cycle requires the development of a partitioning scheme to separate Am from the lanthanides. The Am/lanthanide separation is challenging since all the metals are normally trivalent and have similar ionic radii. Oxidation of Am to higher oxidation states is one option to achieve such a separation. Hexavalent Am has now been routinely prepared in our laboratory in strongly acidic solution using sodium bismuthate as the oxidant, and then extracted into diamylamylphosphonate/dodecane solution. Here, we have characterized this phosphonate-containing solvent with regard to the extraction of Am, the lanthanides, Cm, other fission product, and/or inert constituents expected in dissolved nuclear fuel. Additionally, the effects of irradiation on dispersion numbers and the phosphonate concentration were investigated.

The Partitioning of Americium and the Lanthanides Using Tetrabutyldiglycolamide (TBDGA) in Octanol and in Ionic Liquid Solution

Separations among the lanthanides and the separation of Am from the lanthanides remain challenging, and research in this area continues to expand. The separation of adjacent lanthanides is of interest to high-tech industries because individual lanthanides have specialized uses and are in short supply. In nuclear fuel cycle applications Am would be incorporated into fast-reactor fuels, yet the lanthanides are not desired. In this work, the diglycolamide N,N,N′,N′-tetrabutyldiglycolamide (TBDGA) was investigated as a ligand for lanthanide and Am solvent extraction in both molecular and room-temperature-ionic-liquid (RTIL) diluents.The RTIL [C4MIM][Tf2N] showed very high extraction efficiency for these trivalent metals from low nitric acid concentrations, while the molecular diluent 1-octanol showed high extraction efficiency at high acid concentrations. This was attributed to the extraction of ionic nitrate complexes by the RTIL, whereas 1-octanol extracted neutral nitrate complexes. TBDGA in RTIL did not provide adequate separation factors for Am/lanthanide partitioning, but 1-octanol did show reasonable separation possibilities. Lanthanide intergroup separations appeared to be feasible in both diluents, but with higher separation factors from 1-octanol.