P. N. Pathak

N,N,N′,N′‐Tetraoctyl Diglycolamide (TODGA): A Promising Extractant for Actinide‐Partitioning from High‐Level Waste (HLW)

Abstract N,N,N′,N′‐Tetraoctyl diglycolamide (TODGA) has been evaluated as an extractant for the partitioning of minor actinides from simulated high level nuclear waste solutions. Acid uptake studies suggested that TODGA is more basic (KH: 4.1) as compared to TRUEX and DIAMEX solvents viz. CMPO (KH: 2.0) and DMDBTDMA (KH: 0.32), respectively. TODGA molecules display a tendency toward aggregation in n‐dodecane at lower acidities. Effect of diluent on the distribution behavior of Am(III) was studied employing diluents of different dielectric constants. N,N‐dialkyl amides with different alkyl groups viz. dibutyl decanamide (DBDA), di(2‐ethylhexyl) acetamide (D2EHAA), di(2‐ethylhexyl) propionamide (D2EHPRA), di(2‐ethylhexyl) isobutyramide (D2EHIBA), dihexyl octanamide (DHOA) and dihexyl decanamide (DHDA) were evaluated as phase modifiers. Distribution behavior of various metal ions viz. Am(III), Pu(IV), U(VI), Eu(III), Fe(III), Sr(II) and Cs(I) was studied from pure nitric acid solution as well as from simulated high level waste solution.

Extraction of actinides using N, N, N′, N′-tetraoctyl diglycolamide (TODGA): a thermodynamic study

Summary The effect of temperature on the extraction behaviour of Am(III), Pu(IV) and U(VI) from nitric acid medium was studied employing N,N,N′,N′-tetraoctyl diglycolamide (TODGA) in n-dodecane. The two-phase equilibrium constants (log K′ex) were calculated and compared with those of other extractants proposed for actinide partitioning, viz. octyl-(phenyl)-N,N-diisobutylcarbamoylmethyl phosphine oxide (CMPO) and N,N,N′,N′-dimethyl dibutyl tetradecyl malonamide (DMDBTDMA). Thermodynamic parameters, viz. ΔG, ΔH and ΔS for the extraction of actinides by TODGA were also compared with those of CMPO and DMDBTDMA. These studies indicate that the extraction processes of Am(III) and U(VI) are enthalpy driven whereas entropy factor counteracts the extraction. However, in the case of Pu(IV), the extraction process is enthalpy as well as entropy favoured. Role of diluent on the loading of Nd(III) in 0.1 M TODGA has also been investigated.

Demonstration of T2EHDGA Based Process for Actinide Partitioning Part II: Counter-Current Extraction Studies

Abstract Counter-current mixer-settler studies for actinide partitioning were carried using N,N,N′,N′-tetra-2-ethylhexyl diglycolamide (T2EHDGA) as the extractant. The feed solution was Simulated High Level Waste of Pressurized Heavy Water Reactor (PHWR-SHLW) origin spiked with 241Am, 244Cm, 152Eu, 137Cs, 85,89Sr, 59Fe, 106Ru, 109Pd, 95Zr, and 99Mo tracers. The organic stream was 0.1 M T2EHDGA + 5% isodecanol in n-dodecane. Extraction, scrubbing, and stripping experiments were performed by maintaining an organic to aqueous phase ratio of 1. More than 99.9% of the trivalent actinides and lanthanides were extracted in four stages, and the decontamination factors (D.F.) values were >103 obtained for most fission products. The co-extraction of Zr and Pd was prevented by the addition of oxalic acid and N-(2-hydroxyethyl)-ethylenediamine-triacetic acid (HEDTA) into the feed solution. However, ∼20% Ru and 10% Mo was extracted into the organic phase, which was successfully scrubbed using a mixture of 0.2 M oxalic acid and 0.1 M HEDTA in 5 M HNO3. Finally, the extracted actinides and lanthanides were quantitatively stripped with 0.2 M HNO3. Raffinate of the extraction cycle was found to be free from any alpha activity.

Separation Studies of Uranium and Thorium Using Di-2-Ethylhexyl Isobutyramide (D2EHIBA)

The extraction behavior of di-2-ethylhexyl isobutyramide (D2EHIBA) in dodecane medium for U(VI), Th(IV), and fission products such as Zr, Ce, Eu, and Cs, and the structural material Fe, has been investigated over a wide range of nitric acid concentrations. It has been observed that whereas D U varies from < 10−3 (pH 2.0) to 4.4 (6 M HNO3) with 1 M ligand, the corresponding D Th values are 1.5 × 10−3 and 4 × 10−2. In the presence of 250 g/L of Th, D U values are 8.6 (pH 2.0) and 2.2 (6 M HNO3). D2EHIBA has been found to be a promising extractant of trace concentrations of U in the presence of macro amounts of Th. The extraction of fission products and Fe is found to be negligible. D2EHIBA is found to extract nitric acid predominantly as a 1:1 species (K H = 0.156 ± 0.048). U(VI) is extracted as a disolvate UO2(NO3)·2D2EHIBA (K ex = 0.87 ± 0.08).

Novel polymer inclusion membrane containing a macrocyclic ionophore for selective removal of strontium from nuclear waste solution

A new polymer inclusion membrane (PIM) containing cellulose triacetate (CTA) as the monomer, di-tert-butylcyclohexano-18-crown-6 (DtBuCH18C6) as the carrier and 2-nitrophenyl octyl ether (NPOE) as the plasticizer was developed for the selective transport of Sr2+ from aqueous nitrate medium. Studies with crown ether concentration variation have indicated a linear dependence on the permeability coefficient (P) suggesting a diffusion mechanism for ion transport. Effects of membrane thickness, nature of plasticizer and plasticizer concentration on the transport of Sr2+ were studied. The effect of feed acidity was also investigated for a possible application in the nuclear waste solution. Selective Sr2+ transport was observed in a synthetic waste solution containing metal ions such as UO22+, MoO22+, Zr4+, Ce3+, Nd3+, Ru3+, Pd2+, Ba2+ and Cs+, etc., in 1 M HNO3 and 2 M NaNO3. Greater than 70% transport of Sr2+ was observed in 24 h while most of the other metal ions were negligibly transported (<0.01%).

EVALUATION OF DI(2-ETHYLHEXYL)ISOBUTYRAMIDE (D2EHIBA) AS A PROCESS EXTRACTANT FOR THE RECOVERY OF 233U FROM IRRADIATED Th

ABSTRACT Extraction behaviour of U(VI) and Th(IV) has been studied employing 0.5 M D2EHIBA (di(2-ethylhexyl)isobutyramide) in n-dodecane as a function of nitric acid concentration. An improved separation of U from Th could be achieved in the presence of fluoride ion. Use of 1 M Al(NO3)3 in 2 M HNO3 has been suggested as the scrubbing solution. Various physico-chemical properties like density, viscosity, interfacial tension (IFT) and limiting organic concentration (LOC) of Th/U have been measured to evaluate the use of D2EHIBA as a process solvent. Degradation behaviour (acid hydrolysis as well as gamma radiolysis) of D2EHIBA under process conditions has been investigated using potentiometry as well as distribution studies.

Radiolytic Degradation Studies on N,N‐dihexyloctanamide (DHOA) under PUREX Process Conditions

The radiolytic stability of DHOA, a high molecular weight N,N‐dialkyl amide has been investigated to evaluate its performance under PUREX process conditions vis–a–vis TBP. Gas chromatographic studies revealed the presence of caprylic acid, dihexylamine and dihexylketone in irradiated DHOA. Batch distribution studies of Pu, U, and fission products (144Ce, 103,106Ru, and 137Cs) were carried out using the irradiated samples of 1.1 M DHOA and TBP in n‐dodecane, which showed significant retention of Pu, U, and fission products in the irradiated TBP as compared to that of DHOA even after successive contacts with the stripping solutions. Typically at 60 M Rad dose, the Pu content for DHOA was 1.4 mg/L after three contacts with 0.5 M HNO3, and that for TBP was ∼24 mg/L. White precipitate was observed at the interface during the stripping of Pu (with 0.5 M HNO3) from the loaded irradiated TBP phase. The DHOA system, on the other hand showed no such problem during the stripping cycle but there was an increase in the density and viscosity for the irradiated DHOA.

Separation of uranium and thorium using tris(2-ethylhexyl) phosphate as extractant

The distribution behavior of uranium and thorium has been investigated in a biphasic system of different aqueous nitric acid concentrations and a solution of tris(2-ethylhexyl) phosphate (TEHP) inn-dodecane at 25°C. The effect of different uranium and thorium concentrations in the aqueous phase on the extraction of these metal ions is evaluated. These results indicate that TEHP is a better choice than tri-n-butyl phosphate (TBP) for the separation of233U from the irradiated thorium matrix.

N,N-Dialkyl amides as extractants for spent fuel reprocessing: an overview

Reprocessing of spent nuclear fuel is vital for the long-term global nuclear power growth and is the major motivation for developing novel separation schemes. Conventionally, PUREX and THOREX processes have been proposed for the reprocessing of U and Th based spent fuels employing tri-n-butyl phosphate (TBP) as extractant. However, based on the experiences gained over last five–six decades on the reprocessing of spent fuels, some major drawbacks of TBP have been identified. Evaluation of alternative extractants is, therefore, desirable which can overcome at least some of these problems. Extensive studies have been carried out on the evaluation of N,N-dialkyl amides as extractants in the back-end of the nuclear fuel cycle for addressing the issues related to the reprocessing of U and Th based spent fuels. Under advanced fuel cycle scenario, efforts are also being made by countries with a developed nuclear technological base to provide safe nuclear power to other countries and to minimize proliferation concerns worldwide. This paper presents an overview of studies carried out in our laboratory on different aspects of reprocessing of U and Th based spent fuels employing N,N-dialkyl amides as extractants.

Distribution studies on Th(IV), U(VI) and Pu(IV) using tri-n-butylphosphate and N,N-dialkyl amides

Summary Distribution studies on U, Pu and Th have been carried out employing tri-n-butyl phosphate (TBP) as extractant to optimise the conditions for the reprocessing of spent fuel of proposed Th based Advanced Heavy Water Reactor (AHWR). Due to their high basicity and organophilicity straight chain N,N-dialkyl amides have been evaluated for their third phase formation behaviour (usually quantified by limiting organic concentration) during the extraction of thorium vis a vis TBP. Dihexyl derivatives of octanamide (DHOA) and decanamide (DHDA) were found promising in view of higher limiting organic concentration of thorium viz. 33 g and 50 g/L, respectively as compared to 26 g/L for TBP. Extraction profiles of U, Pu and Th have been obtained using these amides and compared with those of TBP. Attempts have also been made to optimise the stripping conditions using aceto-hydroxamic acid (AHA) and acetaldoxime as reductants for the partitioning of Pu from the loaded organic phase.