High-level Waste Solutions Research Articles

Partitioning of the high-level radioactive waste from the purex process by 40% TBF in isoparaffin using magnesium nitrate as a salting-out agent

Centrifugal contactor rig trials of the partitioning flowsheet were carried out using simulated and real high level waste solutions obtained by reprocessing of WWER-100 spent nuclear fuel with burn-up of 47 MW day/t. 40% TBP in Isopar M was used as a solvent with magnesium nitrate as a salting-out agent. Complete removal of Cs, Sr, Zr and Mo to raffinate was achieved. Rare earth elements, Am and Cm were totally concentrated in TPE and REE product. Comparison of the results of this work with previous trials with iron (III) nitrate as a salting-out agent [1] demonstrated that using magnesium nitrate as a salting-out agent resulted in higher decontamination factors of TPE and REE fraction from Cs (~10000) than using iron (III) nitrate (7500).

РЕЗУЛЬТАТЫ ИСПЫТАНИЙ СОРБЦИОННЫХ МАТЕРИАЛОВ ДЛЯ ИЗВЛЕЧЕНИЯ РАДИОНУКЛИДОВ ЦЕЗИЯ ИЗ ЩЕЛОЧНЫХ ВЫСОКОАКТИВНЫХ ОТХОДОВ ФЕДЕРАЛЬНОГО ГОСУДАРСТВЕННОГО УНИТАРНОГО ПРЕДПРИЯТИЯ «ПО “МАЯК”»

The research results are presented of sorption extraction of cesium from alkaline high-level waste solutions stored in Mayak tanks using the resorcinol-formaldehyde resin RFS-i and modified nickel ferrocyanide-based inorganic sorbent Fersal. It is shown that the overall sorption performance of the Fersal exceeds that of the RFS-i. However, unlike the Fersal, the RFS-i remains stable during repeated cycles. A preliminary conclusion has been drawn on the prospects of use of both sorption materials for decontamination of Mayak real alkaline HLW.

Extraction of Cesium, Strontium, and Stable Simulated HLW Components with Substituted Crown Ethers in New Fluorinated Diluents

ABSTRACT The extraction of cesium, strontium, and a number of stable components of simulated high-level waste solutions from nitric acid media with solutions of crown ether derivatives in new fluorinated diluents was studied. Based on the data on the solubility of the extractants into the aqueous phase and the physicochemical properties of the diluents used, the most promising extraction systems were chosen: crown ethers 4,4‘(5’)-di-tert-butyldibenzo-18-crown-6 (L1) and 4,4‘(5’)-di-tert-butyldicyclohexyl-18-crown-6 (L2) in bis(2,2,3,3-tetrafluoropropyl) carbonate (BK-1) and bis(2,2,3,3-tetrafluoropropoxy)methane (FN-1). For these systems, the extraction of a number of stable components and their subsequent stripping with water and solutions of chelating agents were studied. Lead was found to be co-extracted to the greatest extent. To a lesser extent, barium and calcium cations are co-extracted with strontium, and rubidium and potassium cations with cesium. Lead, barium, and silver are the most difficult to strip; however, the problem of lead stripping can be solved by using DTPA or citric acid solutions neutralized with ammonia to weakly alkaline (pH 8) media.

Precise recognition of palladium through interlaminar chelation in a covalent organic framework

<h2>Summary</h2> Palladium isotopes as fission products in used nuclear fuel represent precious alternative resources besides its natural reserves and therefore have a high extraction value. The practical use of currently proposed solvent extraction is challenging, originating from the limited stability and separation selectivity of the extractants under harsh reprocessing conditions. Herein, we present an interlayer synergistic binding strategy, a metal-recognition manner deviating from chelation within a single ligand molecule in solvent extraction, for selective palladium chelation in a covalent organic framework. The experimental, structural, and theoretical analyses corroborate that the enol-to-keto tautomerization leads to selective synergistic chelation of Pd²⁺ instead of other undesired metal ions, where two oxygen donors from adjacent layers and two free nitrate ions work together in a planar tetracoordination model. Fast adsorption kinetics, high adsorption capacity, and one-round enrichment of Pd²⁺ from the simulated high-level nuclear waste solution are unprecedentedly achieved in the dynamic breakthrough experiment.

Extraction chromatography based separation of zirconium(IV) from simulated high-level liquid waste using N,N-di-octyl-2-hydroxyacetamide impregnated amberlite XAD-7 resin

Selective separation of zirconium(IV), from the simulated high-level liquid waste solution containing uranium(VI) and trivalent lanthanides, during reprocessing of fast reactor metal fuel was studied by extraction chromatography using N,N-di-octyl-2-hydroxyacetamide (DOHyA) impregnated on to Amberlite XAD-7 resin. The extraction behavior of Zr(IV), Nd(III) and U(VI) with varying concentrations of nitric acid was explored using a solvent impregnated resin with different compositions of DOHyA in XAD-7 [10, 20, 30, 40 and 50 (w/w) %]. The resin exhibited faster kinetics with more than 99 % recovery of Zr(IV) and with an apparent Langmuir adsorption capacity of ∼54 mg.g−1of Zr(IV) on to the resin phase. Quantitative (> 99%) and selective extraction of Zr(IV) with minimum loading of other trivalents and uranium was observed during column chromatography. Results of the study showed that Zr(IV) can be selectively separated in presence of trivalent lanthanides and U(VI).

Extraction and aggregation behaviour of Zr(IV) in diglycolamide solvents during the treatment of high-level liquid waste solution arising from metallic fuel reprocessing

Zirconium was identified as a troublesome fission product interfering with the separation of trivalent actinides from high-level liquid waste (HLLW) solution. Since zirconium was used as a fuel component in metallic fuels, the amount of Zr(IV) present in HLLW arising from reprocessing of metallic fuel was nearly ten times higher than the conventional HLLW solutions. In the present study, the diglycolamide extractants proposed for the separation of trivalent actinides from HLLW were employed for the extraction of metal ions from nitric acid medium and fast reactor simulated high level liquid waste (FR-SHLLW) having the Zr(IV) concentrations varied from 0.006 M to 0.09 M. Dynamic light scattering study was performed on the organic phase obtained after extraction of zirconium to analyze the aggregation of extractants. The solvent system consisting of 0.1 M TODGA (N,N,N’,N’-tetraoctyldiglycolamide) along with 0.25 M HDEHP (bis(-2-ethylhexyl)phosphoric acid) in n-dodecane showed minimal aggregation tendency and higher loading capacity of Zr(IV) and was selected for further extraction of trivalent metal ions from FR-SHLLW. To minimize the extraction of Zr(IV) in organic phase, the effect of aqueous soluble complexing agent namely trans-1,2-diaminocyclo-hexane-N,N,N’N’-tetraaceticacid (CyDTA) was also studied during the extraction of trivalent lanthanides.

Selective removal of Zr(IV) from simulated High-Level liquid waste of metallic fuel reprocessing using hydroxyacetamide Extractant: Insights from solvent extraction and density functional theory computations

A liquid-liquid extraction method for the separation of Zr(IV) from simulated high-level liquid waste (SHLLW) solutions using N,N-di-octyl-2-hydroxyacetamide (DOHyA) was studied. The method was found to be applicable for the recovery of Zr(IV) from conventional SHLLW solutions as well as the SHLLW solutions from metallic fuel reprocessing streams containing Zr(IV) up to 8 g/L. DOHyA showed quantitative extraction of Zr(IV) from all the simulated waste solutions, along with co-extraction of other fission products and Am(III). Excluding Zr(IV) and Mo(VI), all the other co-extracted metal ions were efficiently scrubbed from the loaded DOHyA solvent using 0.5 M or 1 M nitric acid. The selective stripping of Zr(IV) that remained in the solvent phase was subsequently accomplished using 0.2 M oxalic acid solution. The complexation behaviour of Zr(IV) with DOHyA was investigated using Density Functional Theory (DFT) calculations, which divulged the geometry and stability of Zr-DOHyA-HNO3 and Zr(C2O4)2 complexes formed under the extraction and stripping conditions. With the aid of quantum theory of atoms in molecules, natural bond orbital and non-covalent interaction analyses, the rationale behind the selective removal of Zr(IV) from SHLLW using DOHyA extractant was clearly unravelled.

Exploring novel functionality for efficient extraction of UO22+ and Th4+in ionic liquid: Mechanism, speciation, selectivity, stability and stripping

Novel ligand functionality, 5-bromo-N1, N3-bis(4,6-dimethylpyridin-2-yl)isophthalamide, was explored for highly efficient and selective separation of UO22+ and Th4+ in pyridinium based ionic liquid. The comprehensive investigation revealed that, both in ionic liquid as well as in molecular diluents the extraction predominantly proceeded via ‘solvation’ mechanism. However, the species involved in extraction were UO2(NO3)2, L and Th(NO3)4, L in ionic liquid, while UO2(NO3)2, 2 L and Th(NO3)4, 3 L in molecular diluent. The highly viscous ionic liquid resulted the mass transfer slower during extraction. In ionic liquid based system, the extraction processes were endothermic, spontaneous and entropy driven. Ionic liquid based solvent systems exhibited high radiation stability, which was attributed to the formation of resonance-stabilized aromatic cation as well as the restricted movement of radicals in viscous ionic liquid. The aqueous solution of sodium carbonate was found to be effective for the stripping of UO22+, while oxalic acid solution was found to be effective for Th4+. The solvent systems were used for processing of simulated high level waste solutions of a Research Reactor (RR) and Fast Breeder Reactor (FBR) origin.

Studies on the separation of Am(III) from trivalent lanthanides in high-level waste solution using modifier-free solvents and aqueous soluble bis-1,2,4-triazines

The selective partitioning of trivalent actinides (An(III)) from high-level liquid waste (HLLW) is considered as an essential step in the management of spent nuclear fuel. The i-SANEX process for the separation of An(III) from the co-existing lanthanides (Ln(III)) in 4 M nitric acid solution and simulated HLLW was studied using a couple of symmetrical diglycolamides (TODGA and TDDGA) and an unsymmetrical diglycolamide (D3DODGA) in n-dodecane as the extraction phase. The selective stripping of Am(III) from the loaded organic phase containing trivalent Ln(III) was investigated using aqueous soluble bis-1,2,4-triazine derivatives such as SO3-Ph-BTP, SO3-Ph-BTBP and SO3-Ph-BTPhen in dilute nitric acid solution. The results revealed that the SF of Eu(III) over Am(III) decreased with increase in the concentration of nitric acid in all cases and SF decreased in the order SO3-Ph-BTP > SO3-Ph-BTBP > SO3-Ph-BTPhen. The co-stripping of lighter lanthanides (La, Ce, Pr, Nd) was also observed during the recovery of Am(III).

Optimization of the Solidification Method of High-Level Waste for Increasing the Thermal Stability of the Magnesium Potassium Phosphate Compound

The key task in the solidification of high-level waste (HLW) into a magnesium potassium phosphate (MPP) compound is the immobilization of mobile cesium isotopes, the activity of which provides the main contribution to the total HLW activity. In addition, the obtained compound containing heat-generating radionuclides can be significantly heated, which increases the necessity of its thermal stability. The current work is aimed at assessing the impact of various methodological approaches to HLW solidification on the thermal stability of the MPP compound, which is evaluated by the mechanical strength of the compound and its resistance to cesium leaching. High-salt surrogate HLW solution (S-HLW) used in the investigation was prepared for solidification by adding sorbents of various types binding at least 93% of 137Cs: ferrocyanide K-Ni (FKN), natural zeolite (NZ), synthetic zeolite Na-mordenite (MOR), and silicotungstic acid (STA). Prepared S-HLW was solidified into the MPP compound. Wollastonite (W) and NZ as fillers were added to the compound composition in the case of using FKN and STA, respectively. It was found that heat treatment up to 450 °C of the compound containing FKN and W (MPP-FKN-W) almost did not affect its compressive strength (about 12–19 МPa), and it led to a decrease of high compressive strength (40–50 MPa) of the compounds containing NZ, MOR, and STA (MPP-NZ, MPP-MOR, and MPP-STA-NZ, respectively) by an average of 2–3 times. It was shown that the differential leaching rate of 137Cs on the 28th day from MPP-FKN-W after heating to 250 °C was 5.3 × 10−6 g/(cm2∙day), however, at a higher temperature, it increased by 20 and more times. The differential leaching rate of 137Cs from MPP-NZ, MPP-MOR, and MPP-STA-NZ had values of (2.9–11) × 10−5 g/(cm2∙day), while the dependence on the heat treatment temperature of the compound was negligible.

Complexation and bonding studies on [Ru(NO)(H2O)5]3+ with nitrate ions by using density functional theory calculation.

Complexation reactions of ruthenium–nitrosyl complexes in HNO3 solution were investigated by density functional theory (DFT) calculations in order to predict the stability of Ru species in high-level radioactive liquid waste (HLLW) solution. The equilibrium structure of [Ru(NO)(NO3)3(H2O)2] obtained by DFT calculations reproduced the experimental Ru–ligand bond lengths and IR frequencies reported previously. Comparison of the Gibbs energies among the geometrical isomers for [Ru(NO)(NO3)x(H2O)5−x](3−x)+/− revealed that the complexation reactions of the ruthenium–nitrosyl complexes with NO3− proceed via the NO3− coordination to the equatorial plane toward the Ru–NO axis. We also estimated Gibbs energy differences on the stepwise complexation reactions to succeed in reproducing the fraction of Ru–NO species in 6 M HNO3 solution, such as in HLLW, by considering the association energy between the Ru–NO species and the substituting ligands. Electron density analyses of the complexes indicated that the strength of the Ru–ligand coordination bonds depends on the stability of the Ru species and the Ru complex without NO3− at the axial position is more stable than that with NO3−, which might be attributed to the difference in the trans influence between H2O and NO3−. Finally, we demonstrated the complexation kinetics in the reactions x = 1 → x = 2. The present study is expected to enable us to model the precise complexation reactions of platinum-group metals in HNO3 solution.

Characterization of high level nuclear waste solutions from different origin

A simple, simultaneously multi-elemental, ICP-AES based method was optimized to characterize nuclear high level waste solutions from different origin: research reactors, pressurized heavy water reactor, thoria containing waste and 137Cs rich solutions. It involves identification of interference free analytical lines with the evaluation of their analytical performance including sensitivity, detection limit etc. To compare the results of ICP-AES system, ED-XRF technique has been employed and the results compared well indicating the efficacy of the optimized procedure. For analytes like Fe, Cr, Na, Mn and Ca the detection limits are in the range of 0.001–0.009 mg/L while for other analytes (except U and Th) the detection limits are almost on the order of magnitude poorer. For all the analytes the linear dynamic range exists from the detection limit up to 500 mg/L.

Separation of strontium-90 from a highly saline high level liquid waste solution using 4,4′(5′)-[di-tert-butyldicyclohexano]-18-crown-6 + isodecyl alcohol/n-dodecane solvent

Selective and efficient separation of 90Sr from high level waste solution of high salt content will pave the way to manage a problematic waste and also for utilization of recovered 90Sr for societal purpose. In the present work, a process for selective separation of 90Sr is described using 4,4′(5′)-[di-tert-butyldicyclohexano]-18-crown-6 (DtBuCH18C6) dissolved in an industrial diluent system: isodecyl alcohol + n-dodecane. The process solvent is optimized as 0.1 M DtBuCH18C6 + 50% isodecyl alcohol/n-dodecane for satisfactory extraction of strontium at higher nitric acid concentrations (D4-8 M HNO3 = 5–14). Stoichiometry determination by slope analysis method showed 1:1:1 M ratio for Sr2+, DtBuCH18C6 and HNO3 in the extracted complex. Among the inert and active constituents of the waste, extraction of plutonium (IV) (DPu = 0.3) and barium (DBa = 0.7) was observed to some extent, extraction of calcium was almost negligible (DCa = 0.01) while no extraction of americium, cesium, ruthenium, niobium, zirconium and lanthanides was observed. Stripping of the loaded organic was achieved by deionized water, three consecutive contacts each with fresh deionized water results in complete stripping of strontium.

Glass ceramic for the vitrification of high level waste with a high molybdenum content

Vitrification has been selected in France as the immobilization process for high-level waste arising from spent fuel reprocessing. Several high-level waste solutions from the reprocessing of legacy UMo spent fuel, used in gas cooled reactors, have been stored in the Orano La Hague facility in stainless steel tanks since the mid-1960s. A special glass-ceramic formulation has been developed and qualified through lab and pilot testing to meet standard waste acceptance criteria for the final disposal of the UMo waste. These solutions are very rich in molybdenum and phosphorus whose contents make the molten glass quite corrosive and require a high-temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum oxide). Molybdenum is known to be sparingly soluble in conventional borosilicate glass. The formulated glass-ceramic matrix comprises a major vitreous phase containing secondary phase particles less than 100 μm in diameter. These are formed by phase separation and crystallization as the molten glass cools. The physical and microstructural properties of the UMo glass in the solid and liquid states are reported here. Evolution of microstructure as a function of the cooling profile was investigated, given the sensitivity of the crystallization process to the cooling profile. The chemical durability of the UMo glass-ceramic was also studied. The feasibility of this process has been demonstrated in a full-scale pilot facility with inactive surrogate solutions.

Comparison in the aggregation behaviour of amide extractant systems by dynamic light scattering and ATR-FTIR spectroscopy

The aggregation behavior of N,N,N’,N’-tetraoctyldiglycolamide (TODGA) in n-dodecane was studied in the presence of two different phase modifiers, N,N-di-hexyloctanamide (DHOA) and 1-octanol using dynamic light scattering (DLS) spectroscopy. The results were compared with the aggregation behavior of a modifier free extractant system N,N-di-octyl-2-hydroxyacetamide (DOHyA) in n-dodecane. Aggregation studies were done as a function of concentration of nitric acid and Nd (III) ions in the aqueous phase. The size distribution of reverse micellar aggregates revealed different aggregation mechanisms in the presence and absence of phase modifiers. For TODGA-modifier systems, the sizes of reverse micellar aggregates were bigger than DOHyA and among the two phase modifiers studied, use of 1-octanol resulted in the formation of bigger aggregates compared to DHOA. The observations from DLS measurements were co-related with the results from ATR-FTIR spectroscopy for all the three extractant systems. The sizes of reverse micellar aggregates increased with increase in extraction of nitric acid and/or Nd (III) ions and no third phase was visible in all the three systems even under very high loading conditions. The aggregation studies under co-current extraction conditions from fast reactor simulated high level waste solution (FR-SHLLW) showed the formation of smaller reverse micellar aggregates and third phase or crud formation after 3–4 successive contacts for all the three systems.

Quaternary ammonium based task specific ionic liquid for the efficient and selective extraction of neptunium

Abstract: Liquid–liquid extraction of neptunium from aqueous acidic solution using quaternary ammonium based task specific ionic liquid (TSIL) was investigated. The extraction of Np was predominated by the ‘cation exchange’ mechanism via [NpO2·Hpth]+ species for NpO2 2+, while NpO2 + was extracted in ionic liquid as [NpO2·H·Hpth]+. The extraction process was thermodynamically spontaneous while kinetically slower. Na2CO3 as strippant showed quantitative back extraction of neptunium ions from TSIL. TSIL showed excellent radiolytic stability upto 500 kGy gamma exposure. Finally, the TSIL was employed for the processing of simulated high level waste solutions revealing high selectivity of TSIL towards neptunium.

Radium and Thorium Sorption by Monosodium Titanate (MST) and Modified MST (mMST)

To confirm performance assessment inputs for the Saltstone Facility at the Savannah River Site (SRS), a series of experiments were performed to examine the removal of Ra and Th by monosodium titanate (MST) and modified MST (mMST). The solubility of Th in the simulated high-level waste solutions was found to be extremely low; therefore, its removal by MST could not be measured. Radium was rapidly removed from the high-level waste simulant by both MST and mMST.

Comparative study on the radiolytic stability of TBP, DHOA, Cyanex 923 and Cyanex 272 in ionic liquid and molecular diluent for the extraction of thorium

The radiolytic stability of TBP, DHOA, Cyanex 923 and Cyanex 272 in ionic liquid vis a vis molecular diluents have been evaluated for the extraction of thorium. The decrease in DTh value was found to follow the trend of irradiated (Xylene + ligand) > irradiated (ionic liquid + ligand). Ionic liquid was found to act as a sink to protect the ligand functionality from radiation damage. On gamma exposure, the other metal ions were also found to be co-extracted into organic phase from simulated high level waste solution of fast breeder reactor origin consequently reducing the selectivity. The reduction in selectivity on gamma exposure was found to be more in xylene than C8mimNTf2 based systems.

Facilitated transport of tetravalent plutonium from nitric acid medium using a novel extractant – BenzoDODA

Facilitated transport of Pu(IV) from nitric acid medium was studied using a novel extractant, BenzoDODA in supported liquid membrane mode. BenzoDODA is a highly selective ligand for Pu(IV), providing very high decontamination factor from other fission products present in high level waste (HLW) solution (nitric acid medium). Various process parameters such as feed acidity, concentration of BenzoDODA, membrane pore size etc., were optimized to achieve effective transport of Pu(IV). Near quantitative (∼97%) transport of Pu(IV) was observed in ∼360min using 0.1M BenzoDODA as carrier in n-dodecane employing 0.45μm PTFE as membrane support when 0.2M N2H4 in 0.6M HNO3 was used as strippant. Transport studies with diluted HLW of PUREX origin showed near negligible transport of other elements present in the solution when near quantitative transport of Pu(IV) was observed. The maximum permeability coefficient (P) value was found to be 2.50×10−4cm/s at 3.85M HNO3 as feed acidity using 0.1M BenzoDODA as carrier. The membrane was found to be stable over six cycles of operation. Thus, this system can be used for separation of Pu(IV) from acidic process streams.

Electrodriven selective transport of Cs+ using chlorinated cobalt dicarbollide in polymer inclusion membrane: a novel approach for cesium removal from simulated nuclear waste solution.

The work describes a novel and cleaner approach of electrodriven selective transport of Cs from simulated nuclear waste solutions through cellulose tri acetate (CTA)/poly vinyl chloride (PVC) based polymer inclusion membrane. The electrodriven cation transport together with the use of highly Cs+ selective hexachlorinated derivative of cobalt bis dicarbollide, allows to achieve selective separation of Cs+ from high concentration of Na+ and other fission products in nuclear waste solutions. The transport selectivity has been studied using radiotracer technique as well as atomic emission spectroscopic technique. Transport studies using CTA based membrane have been carried out from neutral solution as well as 0.4 M HNO3, while that with PVC based membrane has been carried out from 3 M HNO3. High decontamination factor for Cs+ over Na+ has been obtained in all the cases. Experiment with simulated high level waste solution shows selective transport of Cs+ from most of other fission products also. Significantly fast Cs+ transport rate along with high selectivity is an interesting feature observed in this membrane. The current efficiency for Cs+ transport has been found to be ∼100%. The promising results show the possibility of using this kind of electrodriven membrane transport methods for nuclear waste treatment.