PUREX Research Articles

Evaluation of the remaining spent extraction solvent in vermiculite after leaching tests via PIXE analysis

The spent extraction solvent, primarily comprising tributyl phosphate (TBP) and normal dodecane, is waste product generated by the solvent extraction of plutonium and uranium to develop the plutonium–uranium redox-extraction process. The spent solvent should be appropriately treated for disposal or storage. One treatment option for storing the spent solvent is the adsorption of the liquid into vermiculite. The organic liquid is trapped between the vermiculite layers. Our previous experiments on the adsorption of spent solvent into vermiculite have shown that some parts of the loaded solvent gradually leaked from the vermiculite. Therefore, the elution behavior of the loaded solvent into organic diluents, such as hexane or acetone, were evaluated to investigate the adsorption mechanism and capacity. Some loaded solvents easily leaked into the diluent, whereas some remained inside the particle, even after the leaching test. The difference in behavior might be attributed to different adsorption mechanisms, and the amount of the remaining solvent is the adsorption capacity for safely storing the solvent. This study evaluated the vermiculite adsorption capacity via the remaining solvent after washing with diluents, such as ethanol or distilled water. The remaining solvent was analyzed using particle-induced X-ray emission (PIXE) on phosphorous (P) contained in the TBP. PIXE analysis quantitatively analyzed the trace amounts of residual TBP, and n-dodecane and TBP were suggested to be adsorbed to vermiculite by different mechanisms.

Numerical simulation and experiments study on laser ablation decontamination for strontium and cesium from contaminated 316L stainless steels in spent nuclear fuel reprocessing

Radionuclides, 90Sr and 137Cs in this study, generated during spent fuel reprocessing will contaminate the structural material (316 L stainless steels), affecting the safe decommissioning of nuclear facilities. Laser ablation is a promising surface decontamination technology. In this study, the optimal laser parameters (the laser power of 16 W, the scanning speed of 175 mm/s, and the pitch spacing of 10 μm), used for laser ablation decontamination experiments for contaminated 316 L stainless steel in NaOH and HNO3 containing Sr and Cs which simulated the environments of spent nuclear fuel storage pond and Plutonium Uranium Redox Extraction (PUREX) process, were determined by numerical simulation with COMSOL Multiphysics. Laser induced breakdown spectroscopy (LIBS) was used to evaluate the decontamination performance of laser ablation, and no contaminant elements (Sr and Cs) were detected on the steel surface after laser ablation. Moreover, physicochemical technologies (SEM-EDS, XRD, and XPS) were used to characterize the surface chemical states and morphologies of contaminated steel before and after laser ablation. It was found that on the surface of the contaminated steel after laser ablation, δ-ferrite phase was detected, and the passive film was re-formed.

Feasibility study on reprocessing of HTGR spent fuel by existing PUREX plant and technology

Feasibility of reprocessing of High Temperature Gas-cooled Reactor (HTGR) spent fuel by existing Plutonium Uranium Redox EXtraction (PUREX) plant and technology has been investigated. The spent fuel dissolved solution includes approximately 3 times amount of uranium-235 and 1.5 times amount of protonium because of the 3 times higher burnup compared with that of Light Water Reactor (LWR). Then, the heavy metal of the spent fuel is planned to be diluted to 3.1 times by depleted uranium to satisfy the limitation of Rokkasho Reprocessing Plant (RRP) plant. In the present study, recoverability of uranium and plutonium with the dilution is confirmed by a simulation with a reprocessing process calculation code. Moreover, the case without the dilution from the economic perspective is investigated. As a result, the feasibility is confirmed without the dilution, and it is expected that the reprocessed amount is reduced to 1/3 compared with a diluted case even though the facility should be optimized from the perspective of mass flow and criticality.

Hydroxypyridinone-based stabilization of Np(IV) enabling efficient U/Np/Pu separations in the Adapted PUREX process

The classical process to recover uranium (U) and plutonium (Pu) from used nuclear fuel using tributyl phosphate (TBP), namely the Plutonium Uranium Redox EXtraction (PUREX) process, is complicated by the persistent presence of neptunium (Np) and thus requires extra purification steps. The concept of Adapted PUREX seeks to achieve Np recovery by adjusting the valence of the metal more effectively, thereby controlling its behavior more precisely. This study introduces the use of an aqueous hydroxypyridinone chelator, 3,4,3-LI(1,2-HOPO) (abbreviated as HOPO), to dictate the behavior of Np for recovery and meanwhile simplify cumbersome reprocessing steps. The interactions between Np and HOPO were probed mechanistically by way of absorption spectrophotometry, in conjunction with cyclic voltammetry. UV–Vis-NIR spectra illustrated the reduction of NpO22+ to Np4+, with a fast reaction rate. Cyclic voltammetry revealed quasi-reversible processes between the oxidized and reduced forms of the ligand and its Np complexes. The corresponding heterogeneous rate constants (k0) were estimated from the peak-to-peak separation potentials (ΔEp), at ~ 4 – 35 μm/s for both HOPO and NpHOPO, with scan rates of 0.01 – 0.4 V/s. Meanwhile, the electromotive force (EMF) as well as the change of Gibbs free energy (ΔG) were assessed from the half-wave potential (E1/2), demonstrating the completeness of NpO22+ reduction to Np4+ by HOPO. The cumulative formation constant of the resulting NpHOPO complex (logβ101) was determined by metal competition titration to be 42.0 ± 0.6, corroborating the extraordinarily high affinity of HOPO to tetravalent metal ions. The prowess of valence control by HOPO and the high stability of the formed complex resulted in enhanced separations of Np from U and of Pu from U, with a maximum separation factor of ~7000 for both, nearly 90- and 10300-fold higher, respectively, than the values obtained using conventional PUREX formulae.

Organophosphorus-Modified Lanthanide Nitrates as Potential Actinide Oxide Aerosol Surrogates.

In search of suitable simulants for aerosol uranium waste products from Plutonium Uranium Redox Extraction (PUREX) process burns, a series of lanthanide nitrate hydrates ([Ln(κ2-NO3)3·nH2O]) were dissolved in the presence of tributylphosphate (O═P(O(CH2)3CH3)3) referred to as TBP) in kerosene or triphenylphosphate (O═P(O(C6H5) referred to as TPhP) in acetone. The crystal structure of the TPhP derivatives of the lanthanide nitrate series and uranium nitrate were solved as [Ln(κ2-NO3)3(TPhP)3] (Ln = La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu) and [U(O)2(κ2-NO3)2(TPhP)2] (U), respectively. The lanthanide-TBP, Ln, and U were further characterized using FTIR spectroscopy, 31P NMR spectroscopy, thermogravimetric analysis, and X-ray fluorescence spectroscopy. Further, thermal treatment of the lanthanide-TBP, Ln, and U using a box furnace to mimic pyrolysis conditions was found by PXRD analyses to generate a phosphate phase [LnP3O9 or UP2O7) for all systems. The resultant nuclear waste fire contaminant particulates will impact both aerosol transport and toxicity assessments.

Validity of the Simple Solution Concept for the Aqueous Quaternary System U(IV) Nitrate–U(VI) Nitrate–Nitric Acid–Water

This work focuses on the validity of the simple solution concept of the uranium­(IV) nitrate and uranium­(VI) nitrate aqueous mixtures through different means such as water activity or density measurements using a simple solution approach. Since these two elements are mixed together in the plutonium uranium redox extraction (PUREX) process during the partitioning step, which separates uranium from plutonium, the knowledge of behavior of this mixture will help in the modeling of this process. This will allow a more precise code to be developed and thus better modeling of the different components involved in the PUREX process. Lately, thermodynamic codes are developed taking into account deviation from ideality. These deviations from ideality are estimated thanks to a solution theory. The one that will be used in this paper is the simple solution approach, a simplification of the McKay and Perring equation. The simple behavior of the uranium­(IV) and uranium­(VI) mixture will be checked using different methods. The precision of each method will then be discussed. This will be made through the acquisition of water activity, density, and molality of different solutions of uranium­(IV) nitrate and uranium­(VI) nitrate.

Removal of Cesium from Simulated Spent Fuel Dissolver Liquor

With a resurgence of the nuclear industry’s fortunes, waste management will be an even greater consideration; reducing wastes, better segregation and treatment that lower the impact on waste storage facilities and repositories, in particular geological repositories will help the sustainability of the industry. We at the University of Central Lancashire (UCLan) proposed in a previous publication a sequential chromatographic separation process, Alternative Reprocessing Technology (ART) for Fission Products (FPs) and Minor Actinides (MAs) separation from spent fuel dissolver liquor, as an improvement to/replacement for PUREX (Plutonium Uranium Redox Extraction). This publication addresses the removal of one particular fission product, cesium, and its impact on waste management, and down-stream PUREX operations. Although our proposed process is still in its infancy, its impact on the PUREX process could be significant, with major gains in the separation circuit, waste management of High Level Waste (HLW) and subsequently waste disposal to and design of a geological repository. This paper briefly describes; our concept, some preliminary experimental data and why re-classification of the bulk of HLW (High Level Waste) to Intermediate Level Waste (ILW) would be possible.

Organic solvent extraction of uranium from alkaline nuclear waste

There are proliferation issues with the Plutonium Uranium Redox Extraction process due to the possibility of recovering plutonium. The objective of this research was to evaluate different organic extraction ligands that can remove uranium from the nuclear waste and to determine the most effective organic solvent for extracting uranium only, from alkaline media. The results indicate that Alamine 336 in xylene has zero (0%) extraction capability for surrogate fission products at an optimum extraction time of 15 min. Aliquat 336 in xylene has an extraction percentage of 72% for uranium in 60 min. However, Aliquat 336 in toluene extracted 82% of the uranium from the feed solution after 30 min, decreasing to 76% after 60 min.

Analysis of Venting of a Resin Slurry

A resin slurry venting analysis was conducted to address safety issues associated with over-pressurization of ion exchange columns used in the plutonium uranium redox extraction (PUREX) process at the U. S. Department of Energy's Savannah River Site (SRS). If flow to these columns is inadvertently interrupted, an exothermic runaway reaction could occur between the ion exchange resin and the nitric acid used in the feed stream. This reaction generates significant quantities of noncondensable gases. To prevent the column from rupturing due to pressurization by these gases, rupture disks are installed on the column vent lines. The venting analysis models accelerating rate calorimeter (ARC) tests and data from tests that were performed in a vented test vessel with a rupture disk. The tests showed that the pressure inside the test vessel continued to increase after the rupture disk opened, though at a slower rate than prior to the rupture. The increase in the vessel pressure is modeled as a transient phenomenon associated with expansion of the resin slurry/gas mixture upon rupture of the disk. It is postulated that the maximum pressure at the end of this expansion is limited by energy minimization to approximately 1.5 times the rupture disk burst pressure. The magnitude of this pressure increase is consistent with the measured pressure transients. The results of this analysis demonstrate the need to allow for a margin between the design pressure and the rupture disk burst pressure in similar applications.

Distribution of Fission Products into Tributyl Phosphate under Applied Nuclear Fuel Recycling Conditions

Tributyl phosphate (TBP) is an important industrial extractant used in the Plutonium Uranium Redox Extraction (PUREX) process for recovering uranium and plutonium from used nuclear fuel. Distribution data have been assessed for a variety of fission and corrosion product trace metals at varying uranium concentrations under representative PUREX extraction (3 M HNO3) and stripping (0.1 M HNO3) conditions. As might have been anticipated, the extraction of most trace metals was found to decrease or remain constant with increasing uranium concentration. In contrast, the extraction of some low valence transition metals was found to increase with increasing uranium concentration. The increase in extraction of low valence transition metals may be related to TBP forming reverse micelles instead of recovering uranium as a classical UO2(NO3)2(TBP)2 coordination complex. The low valence transition metals may be being recovered into the cores of the reverse micelles. Also unanticipated was the lack of impact the TBP de...

Synthesis and extraction studies with a rationally designed diamide ligand selective to actinide(iv) pertinent to the plutonium uranium redox extraction process.

A new class of conformationally constrained oxa-bridged tricyclo-dicarboxamide (OTDA) ligand was rationally designed for the selective extraction of tetravalent actinides pertinent to the Plutonium Uranium Redox EXtraction (PUREX) process. Two of the designed diamide ligands were synthesized and extraction studies were performed for Pu(iv) from HNO3 medium. The mechanism of extraction was investigated by studying various parameters such as feed HNO3, NaNO3 and OTDA concentrations. The nature of the extracted species was found to be [Pu(NO3)4(OTDA)]. One of the OTDA ligands was elaborately tested and showed the selective extraction of Pu(iv) and Np(iv) over other actinide species, viz., U(vi), Np(v), Am(iii), lanthanides and fission products contained in a nuclear waste from the PUREX process. DFT calculations predicted the charge density on each of the coordinating 'O' atoms of OTDA supporting its high Pu(iv) selectivity over other ions studied and also provided the energy optimized structure of OTDA and its Pu(iv) complex.

First demonstration of a centrifugal solvent extraction process for minor actinides from a concentrated spent fuel solution

The efficiency of minor actinides (MA) recovery in the diamide extraction (DIAMEX) process has already been demonstrated using high active raffinate (HAR). In view of an industrial application, the next development step aims at the partitioning of MA from high active concentrate (HAC) as feed. The volume reduction reduces the size of the installation, and thereby the amount of waste and the costs of the process. In this work, MA recovery by DIAMEX partitioning using genuine HAC has for the first time been demonstrated. The experiment was successfully carried out in a 16-stage centrifugal extractor battery, installed in the hot cell facilities at ITU, using N, N′-dimethyl- N, N′-dioctyl-2-(2-(hexyloxy)ethyl)-malonamide (DMDOHEMA) as extractant. In order to produce a representative HAC, a concentration–denitration process was applied to a genuine HAR solution produced by small scale plutonium uranium redox extraction (PUREX) reprocessing of a mixed oxide (MOX) fuel solution. In the DIAMEX process, five extraction stages were sufficient to achieve feed decontamination factors above 10,000 for Am and Cm. Co-extraction of molybdenum, zirconium and palladium was prevented using oxalic acid and HEDTA scrubbing. The back-extraction proved to be very efficient, yielding in four stages recoveries for Am and Cm in the product fraction above 99.7 and 99.9%, respectively, which can be further improved by process flowsheet optimisation. This work is a major contribution in the field of partitioning and an important step towards the industrial implementation of MA partitioning.

Partitioning of minor actinides from HLLW using the DIAMEX process.Part 2 - ´´Hot´´ continuous counter-current experiment

Among several processes proposed world-wide, the French DIAMEX (DIAMide EXtraction) process seems to be very efficient for the removal of Minor Actinides (MA) from genuine High Level Liquid Waste (HLLW). The MA are in this process directly extracted from the PUREX (Plutonium Uranium Redox EXtraction) raffinate together with fission lanthanides using the completely combustible diamide extractant. In this work a hot demonstration of the DIAMEX process using genuine high-level PUREX raffinate is reported. The continuous counter-current experiment was carried out in a 16 stage centrifugal extractor battery, installed in a hot cell. In order to produce a representative HLLW a PUREX process was applied on dissolved fuel using the same equipment. In the DIAMEX process up to 6 extraction stages were sufficient to achieve feed decontamination factors between 100 and 230 for lanthanides and above 300 for minor actinides. Co-extraction of molybdenum and zirconium were efficiently prevented using oxalic acid scrubbing. The back extraction proved to be very efficient, yielding in 4 stages more than 99.9% recovery of both the lanthanides and the actinides. Co-extracted ruthenium, technetium, palladium and neptunium are less efficiently back-extracted requiring further process development.

How ‘CIVEX’ and ‘PUREX’ differ

How ‘CIVEX’ and ‘PUREX’ differ