Separation Studies on Long Lived Radionuclides Using Novel Extractants

A

Thesis submitted to the UNIVERSITY OF MUMBAI

for the Degree of

DOCTOR OF PHILOSOPHY

In

CHEMISTRY

By

SERAJ AHMAD ANSARI

Under the guidance of

Prof. V.K. MANCHANDA

Radiochemistry Division Bhabha Atomic Research Centre

Mumbai 400 085

December 2007

i

STATEMENT BY THE CANDIDATE UNDER ORDINANCE 770

As required by the University Ordinance 770, I wish to state that the work embodied in this thesis entitled Separation Studies on Long Lived Radionuclides Using Novel Extractants forms my own contribution to the research work carried out under the guidance of Prof. V.K. Manchanda, at the Bhabha Atomic Research Centre, Mumbai 400 0085. This work has not been submitted previously for any other degree of either Mumbai University or any other University. Whenever references have been made to previous works of others, it has been clearly indicated as such and included in the Bibliography.

(Ansari Seraj Ahmad) Candidate I hereby, certify that the above statement is correct.

(Prof. V. K. Manchanda) Research Guide

Contents

ii

CONTENTS

Acknowledgements vii

Synopsis of the thesis viii

1. GENERAL INTRODUCTION 1-31

1.1. Nuclear Energy 1

1.2. Nuclear Fuel Cycle 2 1.2.1. Waste from Front End of Fuel Cycle 3

1.2.2. Waste from Back End of Fuel Cycle 4

1.3. Classification of Radioactive Waste 5 1.3.1. Low Level Waste 6 1.3.2. Intermediate Level Waste 6

1.3.3. High Level Waste 6

1.4. Impact of Radionuclides on Environment 7

1.5. Chemistry of Actinides 8 1.5.1. History 8 1.5.2. Electronic Configuration 9

1.5.3. Solution Chemistry of Actinides 9 1.5.3.1. Oxidation States 10

1.5.3.2. Disproportionation Reactions 12

1.5.3.3. Hydrolysis and Polymerization 13

1.5.3.4. Complexation of Actinides 14

1.5.3.5. Absorption Spectra 15

1.6. Separation of Metal Ions 16

1.7. Criteria for Selection of Extractants 18

1.8. Reprocessing of Spent Fuel 19 1.8.1. PUREX Process 19

1.9. Actinide Partitioning 20 1.9.1. TRUEX Process 21

1.9.2. TRPO Process 23

Contents

iii

1.9.3. DIDPA Process 23 1.9.4. DIAMEX Process 24

1.10. DIGLYCOLAMIDES: A Class of Promising Extractants for Actinide Partitioning

25

1.10.1. Main Features of TODGA 26

1.11. Scope of the Thesis 27

1.12. References 28

2. EXPERIMENTAL 32-55

2.1. Synthesis of N,N,N,N-Tetraoctyl Diglycolamide 32

2.2. Characterization of Tetraoctyl Diglycolamide 34

2.3. Synthesis of Malonamide Functionalized Polymer 35

2.4. Characterization of Malonamide Grafted Polymer 36

2.5. Radiotracers (Separation and Purification) 38 2.5.1. Uranium-233 38 2.5.2. Thorium-234 38 2.5.3. Neptunium-239 39

2.5.4. Iron-59 39 2.5.5. Other Radiotracers 40

2.6. Preparation of Simulated High Level Waste 40

2.7. Methods and Equipments 41

2.7.1. Solvent Extraction Studies 41 2.7.2. Mixer-Settler Studies 42 2.7.3. Extraction Chromatography Studies 43 2.7.4. Membrane Studies 44

2.7.5. Hollow Fibre Membrane 45 2.7.6. Other Equipments 47

2.8. Analytical Instruments / Techniques 47 2.8.1. Liquid Scintillation Counter 48

2.8.2. NaI(Tl) Scintillation Counter 49 2.8.3. Surface Barrier Detector 49 2.8.4. High Purity Germanium Detector 51

Contents

iv

2.8.5. Estimation of Uranium 51 2.8.5.1. Spectrophotometry 51

2.8.5.2. Davis Gray Titration 52

2.8.6. Estimation of Thorium 52 2.8.6.1. Spectrophotometry 52

2.8.6.2. Complexometric Titration 53

2.8.7. Estimation of Neodymium 53

2.8.7.1. Spectrophotometry 53

2.8.7.2. Complexometric Titration 53

2.9. References 54

3. N,N,N,N-TETRAOCTYL DIGLYCOLAMIDE: A PROMISING EXTRACTANT FOR THE PARTITIONING OF ACTINIDES FROM HIGH LEVEL WASTE

56-91

3.1. Introduction 56

3.2. Evaluation of Extractants for Actinide Partitioning 57

3.3. Basicity of TODGA 59

3.4. Extraction of Americium by TODGA 60 3.4.1. Effect of Anion 61

3.4.2. Effect of Ligand Concentration 62 3.4.3. Effect of Organic Diluent 65 3.4.4. Kinetics of Extraction 66

3.5. Thermdynamics of Extraction 66

3.5.1. Calculation of Thermodynamic Parameters 67 3.5.2. Effect of Temperature on Distribution of Actinides 70 3.5.3. Thermodynamic Parameters (G, H And S) 71

3.6. Neodymium Loading Studies 74

3.6.1. Evaluation of Phase Modifiers 76

3.7. Extraction of Actinides and Other Metal Ions 78

3.8. Stability of TODGA 82

3.9. Counter-Current Extraction 84 3.9.1. Optimization of Parameters 84

Contents

v

3.9.2. Mixer-Settler Runs 86

3.10. References 88

4. EXTRACTION CHROMATOGRAPHIC STUDIES ON ACTINIDES AND OTHER METAL IONS USING N,N,N,N-TETRAOCTYL DIGLYCOLAMIDE AS THE STATIONARY PHASE

92-113

4.1. Introduction 92

4.2. Preparation of Chromatographic Resin 93

4.3. Batch Studies 95 4.3.1. Evaluation of Resin Materials 95 4.3.2. Kinetics of Extraction of Americium 96

4.3.3. Sorption of Metal Ions on TODGA/Chromosorb-W 97 4.3.4. Sorption of Am(III) Under Loading Conditions 100 4.3.5. Sorption of Am(III) from Nitrate and Sulphate Media 102 4.3.6. Sorption of Metal ions from Synthetic Waste Solution 103

4.4. Column Studies 104 4.4.1. Performance of Chromatography Column 104 4.4.2. Column Breakthrough for Am(III) 107 4.4.3. Column Elution Studies 108

4.4.4. Reusability of Column 110

4.5. References 111

5. SORPTION BEHAVIOUR OF ACTINIDES ON N,N-DIMETHYL-N,N-DIBUTYL MALONAMIDE GRAFTED POLYMER

114-135

5.1. Introduction 114

5.2. Sorption Kinetics for Actinides 115

5.3. Uranium Sorption Studies 118

5.3.1. Sorption Isotherms 118 5.3.2. Sorption Mechanism 123

5.4. Effect of Feed Acidity on Metal Ion Sorption 125

5.5. Desorption Studies 127

Contents

vi

5.6. Analytical Applications 128 5.6.1. Metal Loading Capacity 129 5.6.2. Tolerance of Metal ions on Sorption of Uranium 130

5.6.3. Column Separation of Am, Pu and U 130 5.6.4. Pre-concentration of Uranium and Thorium 133

5.7. References 133

6. TRANSPORT BEHAVIOUR OF LONG LIVED RADIONUCLIDES ACROSS LIQUID MEMBRANES USING N,N,N,N- TETRAOCTYL DIGLYCOLAMIDE AS THE CARRIER

136-168

6.1. Introduction 136

6.2. Theory of Facilitated Transport 137 6.2.1. Distribution Equilibria at Aqueous Membrane Interface 138 6.2.2. Flux Equations for Permeation 139

6.3. Transport of Americium 142

6.3.1. Effect of Membrane Soaking Time 142 6.3.2. Effect of Feed Acidity 143 6.3.3. Effect of Carrier Concentration 145 6.3.4. Effect of Strippant 146

6.3.5. Effect of Nitrate ion Concentration 147

6.4. Transport of Metal ions from Nitric Acid 149

6.5. Transport of Metal ions from SHLW 153

6.6. Stability of Liquid Membrane 156

6.7. Hollow Fibre Liquid Membrane Studies 159 6.7.1. Permeation of Metal Ions across HFSLM 159

6.7.1.1. Transport of Neodymium from HNO3 Solution 160

6.7.1.2. Transport of Americium from SHLW 164

6.7.2. Stability of Liquid Membrane in HFSLM 165

6.8. References 166

Summary and Conclusions 169 Statement Under Ordinance 771 173

vii

ACKNOWLEDGEMENTS

I am deeply indebted to Prof. V.K. Manchanda, Head, Radiochemistry

Division, Bhabha Atomic Research Centre, Mumbai for his invaluable guidance,

critical comments and constant encouragement during the entire course of this study.

I take this opportunity to state that his keen interest and valuable suggestions were of

immense help in improving the quality of work as well as enriching my knowledge.

It is my pleasure to express my sincere thanks to Dr. P.K. Mohapatra, Dr.

P.N. Pathak, Mr. A. Bhattacharyya and Mr. D.R. Prabhu for their active help and

continuous support at all stages of this work. I wish to express my sincere gratitude

to Dr. B.S. Tomar, Dr. M.S. Murali, Mrs. Neetika Rawat, Mr. Sumit Kumar, Ms.

Aishwarya Jain, Mr. R.B. Gujar, Mr. A.S. Kanekar and Mr. D.R. Raut for their

invaluable support and co-operation during the course of this work. I take this

opportunity to thank the technical and administrative staff of Radiochemistry

Division for their immense help during the entire course of this work.

I am thankful to Director, BARC and Director, RC & I Group, BARC for

allowing me to avail all the facilities required for the completion of this work.

Thanks are due to Department of Atomic Energy, Government of India for providing

me the fellowship during the course of this study.

Finally, my family being a constant source of inspiration to me, I take this

opportunity to express my profound gratitude to my beloved family.

viii

SYNOPSIS of the Thesis submitted to the

UNIVERSITY OF MUMBAI for the Degree of

DOCTOR OF PHILOSOPHY IN CHEMISTRY

------------------------------------------------------------------------------------- Title of the Thesis : Separation Studies on Long Lived Radionuclides

Using Novel Extractants Name of the Candidate : Seraj Ahmad F. A. Ansari Name and Designation : Prof. V.K. Manchanda of the Research Guide Head, Radiochemistry Division, Bhabha Atomic Research Centre, Mumbai 400 085 Place of research work : Radiochemistry Division, Bhabha Atomic Research Centre, Mumbai 400 085 Registration Number : BARC - 67 Date of Registration : 26 / 03 / 2004

Signature of the student Signature of the guide

(S. A. Ansari) (Prof. V. K. Manchanda)

Synopsis

ix

Synopsis

Separation Studies on Long Lived Radionuclides Using Novel Extractants

Nuclear energy has been projected as one of the potential sources of energy by

several nations including India. The basic nuclear reaction of neutron induced fission

results in the release of enormous amount of energy. However, due to limited natural

resources of the fissile material (235U), the future nuclear energy program largely

depends upon the availability of the man made fissile materials such as 239Pu and 233U. To sustain nuclear power programme beyond the availability of naturally

occurring 235U, it is imperative to follow the closed fuel cycle option. The closed fuel

cycle emphasizes on recycling of the spent fuel and has been opted by several

countries including India. During reprocessing of the spent fuel, the valuable

plutonium and uranium are recovered by a hydrometallurgical process leaving behind

highly radioactive liquid waste solution referred to as High Level Waste (HLW).

This HLW solution comprises long-lived alpha emitting actinides such as 241Am, 243Am, 245Cm and 237Np (referred as minor actinides) apart from the small amounts of

un-recovered plutonium and uranium as well as beta / gamma emitting fission

products and significant concentrations of structural materials along with process

chemicals. Since the half lives of minor actinides and some of the fission products

range from few hundred to millions of years, HLW poses long term radiological risk

to the environment [1]. The sustainability of the future nuclear energy programme,

therefore, depends upon the effective radioactive waste management which must safe

guard the human health as well as the ecology.

The challenge for the final disposal of HLW is largely due to the radiotoxicity

associated with the minor actinides. At present, the most accepted concept for the

management of HLW is to vitrify it in the glass matrix followed by disposal in deep

geological repositories. Since the half lives of minor actinides concerned range

between a few hundred to millions of years, the surveillance of HLW for such a long

period is economically as well as environmentally daunting task. An alternative /

complimentary concept is the partitioning and transmutation (P&T), which envisages

the complete removal of minor actinides from HLW and their consequent burning in

reactors as mixed oxide fuels [2]. This process would lead to generation of extra

Synopsis

x

energy and at the same time would alleviate the need for long term surveillance of

geological repositories. After partitioning of the actinides along with the long lived

fission products, the residual waste can be vitrified and buried in subsurface

repositories at a much reduced risk and cost. Efforts are being made by radiochemists

/ separation chemists to develop efficient and environmentally benign processes for

the separation of long-lived radionuclides from HLW solution.

For the partitioning of actinides from HLW solution, several processes have

been proposed, viz. TRUEX, DIAMEX, DIDPA and TRPO which employ

octyl(phenyl)-N,N-diisobutyl carbamoyl methyl phosphine oxide (CMPO), N,N-

dimethyl-N,N-dibutyl tetradecyl malonamide (DMDBTDMA), diisodecyl

phosphoric acid (DIDPA) and trialkyl phosphine oxide (TRPO) as the extractants [3].

However, each of the above mentioned processes is associated with certain

limitations. The main drawbacks of the TRUEX process are: (a) the poor back

extraction of Am(III) and Cm(III), and (b) interference due to solvent degradation

products. On the other hand, DIDPA process cannot be applied to the concentrated

HLW solution without denitration which leads to the precipitation of actinides.

Similarly, the TRPO process works only at relatively lower acidity (1M HNO3) and,

therefore, cannot be applied directly to HLW conditions (3-4M HNO3). Though the

completely incinerable DMDBTDMA has been reported to be a promising candidate,

it is a moderate extractant for Am(III) / Cm(III) from HLW solution at acidity 3M

HNO3 [4]. In order to improve the efficiency of diamides towards the forward

extraction of trivalent actinides, several structural modifications of the ligand have

been attempted. Recently, a series of diamide compounds have been synthesized by

introducing different substituents on amide nitrogen or introducing an ether oxygen

into the bridging chain of malonamide [5]. It has been observed that the introduction

of etherial oxygen between the two amide groups (diglycolamide) causes significant

enhancement in the extraction of trivalent actinides / lanthanides. Amongst the

several derivatives of diglycolamide studied, N,N,N,N-tetraoctyl diglycolamide

(TODGA) has been identified as one of the most promising extractants for the

partitioning of trivalent actinides and lanthanides from HLW solutions [6]. Some of

the salient features of TODGA include; (i) large extraction capacity for trivalent

actinides from moderate acidic aqueous solutions, (ii) low concentration of TODGA

Synopsis

xi

(0.1M) to be used, (iii) possibility of complete incineration as the constituent

elements are C, H, N and O, (iv) good radiolytic and hydrolytic stability, and (v) the

ease of synthesis. As TODGA exhibits excellent properties required by an extractant,

it was evaluated for the partitioning of actinides from HLW solution.

The main objective of the present work is to explore the separation of various

radionuclides (actinides / long lived fission products) from structural elements (Fe,

Co, Ni), process chemicals and daughter products of fission products present in

HLW. The present research work includes synthesis and characterization of

extractant / extraction chromatographic material, distribution behaviour of actinides

and other metal ions present in HLW and optimization of experimental parameters

for hollow fibre liquid membrane as well as for mixer-settler runs. Effort has been

made to understand the basic chemistry of TODGA interactions with actinides. An

insight into the sorption behaviour of actinide ions on a novel malonamide grafted

polymer has also been described. The thesis is structured into six chapters for

presentation of the present research work.

CHAPTER-1: GENERAL INTRODUCTION This is the introductory chapter of the thesis that elaborates the importance of the

separation of minor actinides and long-lived fission products from radioactive waste

solutions. The source of these radionuclides and their impact on the environment has

been discussed. The radionuclides which are of major concern are the long lived

alpha emitting radioisotopes which belong to the actinide elements of the periodic

table. The chemistry of actinides is important for their separation and, therefore, the

chemistry of actinides in brief has been presented in this chapter. A brief overview of

the literature reports on the importance and separation of radionuclides by different

extractants has been presented. A brief background of the development of

diglycolamide extractants has been included in this chapter. This chapter also deals

with the aims and objectives of the present work.

CHAPTER- 2: EXPERIMENTAL A general outline about different experimental techniques and instrumentation used

in the present work is given in this chapter. The synthesis, purification and

Synopsis

xii

characterization of TODGA have been described. Synthesis and characterization of a

novel malonamide grafted polymer has also been described. A brief mention about

the various analytical techniques followed is also made in this chapter. For

characterization of materials, techniques like UV-visible absorption spectroscopy,

infrared (IR) spectroscopy and nuclear magnetic resonance (NMR) spectroscopy

were employed. The gamma spectrometry was carried out using NaI(Tl) detector and

HPGe detector, whereas surface barrier detector and liquid scintillation counter were

employed for alpha spectrometry and gross assaying of alpha activity. The basic

principles of these detectors are also described. The preparation and purification of

various radiotracers is included in this chapter. The UV-visible absorption

spectrophotometry was followed for the analysis of Nd, Th and U when their

concentrations were in the range of microgram / mL quantities. The complexometric

titrations carried out for the analysis of various elements such as lanthanides, thorium

and uranium are also described in this chapter.

CHAPTER-3: N,N,N,N-TETRAOCTYL DIGLYCOLAMIDE: A

PROMISING EXTRACTANT FOR THE PARTITIONING OF ACTINIDES FROM HIGH LEVEL WASTE

N,N,N,N-tetraoctyl diglycolamide (TODGA) has been evaluated as an extractant for

the partitioning of minor actinides from radioactive waste solutions [6]. This chapter

deals with the basic solvent extraction chemistry of actinides and fission products

with TODGA. The performance of TODGA for the extraction of actinides has been

compared with those of other extractants proposed for actinide partitioning, viz.

CMPO, TRPO and DMDBTDMA. Acid uptake studies suggested that TODGA is

more basic (KH: 4.1) as compared to CMPO (KH: 2.0) and DMDBTDMA (KH: 0.32).

In order to understand the effect of diluent on the complexation of TODGA with

trivalent actinides the distribution behaviour of Am(III) was studied employing

diluents with different dielectric constants. The effects of complexing anions such as

NO3-, ClO4- and Cl- were investigated to understand the mechanism of extraction for

the metal ions. The thermodynamics of extraction of actinide ions such as Am(III),

Pu(IV) and U(VI) from nitric acid medium by TODGA has also been discussed in

this chapter. The two-phase equilibrium constants and thermodynamic parameters,

Synopsis

xiii

viz. G, H and S for the extraction of actinides have been calculated and

compared with those of CMPO and DMDBTDMA.

One of the important criteria for a good extractant to be used in the solvent

extraction process is the high metal loading capacity in the organic phase. Though

TODGA exhibits high extraction behaviour for trivalent actinides, it forms third

phase at very low metal ion concentration and the limiting organic concentration

(LOC) for neodymium was found to be very low (~0.008M Nd by 0.1M TODGA /

dodecane at 3M HNO3). Third phase formation refers to the phenomenon in which

the organic phase splits into two phases, one is lighter in weight and rich in diluent,

and other is heavier in weight and rich in ligand-metal / ligand-acid complex. Third

phase formation is a natural phenomenon arising out of the incompatibility of the

polar metal solvate species (or acid ligand complex) with the highly non-polar

diluent like dodecane. The third phase is often eliminated by the addition of a

suitable diluent modifier which increases the polarity of diluent thereby increasing

the solubility of metal-ligand complex. In the present work, N,N-dihexyl octanamide

(DHOA) was found to be a promising phase modifier amongst a series of compounds

studied, viz., dibutyl decanamide, di(2-ethylhexyl) acetamide, di(2-ethylhexyl)

propionamide, di(2-ethylhexyl) isobutyramide, dihexyl decanamide, tri-n-butyl

phosphate and 1-decanol. The distribution behaviour of actinides / fission products

has been studied from pure nitric acid solution as well as from synthetic HLW

solution employing 0.1M TODGA + 0.5M DHOA in n-dodecane. This chapter also

reports the applicability of TODGA for the extraction of lanthanides / actinides on

large scale in counter-current extraction using a mixer-settler system.

CHAPTER-4: EXTRACTION CHROMATOGRAPHIC STUDIES ON

ACTINIDES AND OTHER METAL IONS USING N,N,N,N-TETRAOCTYL

DIGLYCOLAMIDE AS THE STATIONARY PHASE In the view of their continuous nature, solvent extraction processes are extensively

employed for plant scale operations for the recovery of metal ions in large scale.

However, the major problem associated with this technique is the generation of large

volume of secondary waste and handling of large volume of inflammable diluents,

particularly when the metal quantities involved are in the grams / milligrams

Synopsis

xiv

quantities. It is, therefore, imperative to look for an alternative technique where the

metal ions can be concentrated in a small volume with minimum generation of

secondary waste. In this context, several techniques like liquid membrane,

magnetically assisted chemical separation (MACS) and extraction chromatography

(EC) are promising alternatives [7-9]. Amongst these techniques EC is rather well

known.

This chapter deals with the preparation of a novel extraction chromatographic

resin impregnated with TODGA and its use to study the sorption behaviour of

actinides / fission products from nitric acid solutions as well as from SHLW solution.

The performance of the present resin has been compared with the resin prepared by

impregnation of other proposed extractants for actinide partitioning such as CMPO,

TRPO and DMDBTDMA. The possibility of the resin material to sorb trace

concentrations of Am(III) from nitric acid solutions containing relatively large

amounts of Nd(III), U(VI), Fe(III) as well as from SHLW solution has also been

reported. In the column chromatographic studies breakthrough capacity of the

column in the presence of macro concentrations of europium and uranium was

investigated. The breakthrough capacity of the column was found to be 20mg of Eu/g

of resin. Elution studies of Am(III) suggested that 0.01M EDTA was effective

amongst different eluents studied.

CHAPTER-5: SORPTION BEHAVIOUR OF ACTINIDE IONS ON

N,N-DIMETHYL-N,N-DIBUTYL MALONAMIDE GRAFTED POLYMER

Solid phase extraction has been increasingly used for the separation of trace as well

as ultra trace amounts of metal ions from complex matrices [10,11]. Chelating

polymers have been frequently used for solid phase extraction of metal ions as they

provide good stability and high sorption capacity. There are two approaches which

are frequently adopted for designing such chelating polymers. The first involves the

physical sorption of chelating ligands on the inert polymeric solid support as

discussed in chapter 4. The other is based on co-valent coupling of the ligands with

the polymer backbone through certain functional groups such as N=N- or CH2-

groups. The latter strategy renders the chromatographic system free from ligand

leaching problem which is often encountered in the former.

Synopsis

xv

Studies on substituted diamide suggested good metal extraction behaviour,

high radiolytic stability and complete incinerability [12]. However, despite these

features, amides do possess inherent limitations such as finite aqueous phase

solubility and third phase formation. In order to overcome these problems, the

synthesis of a novel malonamide grafted polymer was carried out using N,N-

dimethyl-N,N-dibutyl malonamide (DMDBMA) as chelating ligand and Merrifield

polymer as the support backbone. The synthesized polymeric material exhibited

superior binding for hexavalent and tetravalent metal ions such as U(VI) and Pu(IV)

over trivalent metal ions, viz. Am(III) and Pu(III). Various physico-chemical

properties of the polymer like phase adsorption kinetics, metal sorption mechanism

and metal sorption capacity have been studied in the static method. The kinetics for

the adsorption of Am(III), Th(IV) and U(VI) was found to follow the first order

Lagergren rate kinetics. Adsorption of U(VI) on the malonamide grafted polymer

followed the Langmuir adsorption isotherm. The metal sorption capacity for uranium

and thorium by the malonamide functionalized polymer is also reported in this

chapter. Batch extraction studies suggested the possible separation of uranium,

americium and plutonium from each other. The pre-concentration of thorium and

uranium from a large volume of dilute solution employing the grafted resin column is

also reported in this chapter.

CHAPTER-6: TRANSPORT BEHAVIOUR OF LONG LIVED

RADIONUCLIDES ACROSS LIQUID MEMBRANES USING N,N,N,N-

TETRAOCTYL DIGLYCOLAMIDE AS THE CARRIER During the last two decades, the development of selective receptor molecules for

cationic as well as anionic, organic, or inorganic substrates led to their use as carrier

agents for facilitating selective transport through artificial or biological membranes.

Thus, the studies on transport processes were prompted by the design of synthetic

carrier molecules [13]. Liquid membrane transport processes, where the carrier

facilitates selective transportation, have many advantages over solvent extraction.

Liquid membrane processes are being widely employed for the separation of metal

ions involving bulk liquid, supported liquid, or emulsion liquid membranes [14,15].

Facilitated transport of metal ions through liquid membrane has potential

Synopsis

xvi

applications in the nuclear industry such as recovery of metals from

hydrometallurgical leach solutions, treatment and concentration of low level aqueous

waste from reprocessing plants and from waste streams of radiochemical laboratories

engaged in analytical and research activities. This is a fascinating separation

technique because of relatively small inventory of the extractant and low energy

consumption.

This chapter deals with the carrier mediated transport of actinides / fission

products from nitric acid medium across a membrane impregnated with TODGA in

n-dodecane. Microporous PTFE membranes have been used as the polymeric

support. The permeability of transported species through the liquid membrane is

explained in this chapter with the help of various diffusional parameters. Influence of

various parameters, viz. feed acidity, carrier concentration, nature of strippant and

effect of radiation dose on the transport of actinides has been reported. The effect of

macro concentration of neodymium, uranium and iron on the transport of Am(III) has

been illustrated in this chapter. The transport of actinides, fission products and

structural elements from Simulated High Level Waste (SHLW) solution has also

been investigated. The effect of various strippants, namely distilled water, oxalic acid

and buffer solution on the transport of Am(III) has been explored. The membrane

stability was remarkably good when tested over 20 days of continuous operation. The

applicability of membrane separation process on a larger scale has been successfully

demonstrated in a liquid cell contactor (Hollow Fibre Module) for the separation of

lanthanide using TODGA as the extractant.

SUMMARY AND CONCLUSIONS The present thesis describes the separation chemistry of actinides employing

N,N,N,N-tetraoctyl diglycolamide (TODGA) as the extractant. The synthesis,

characterization and interaction of TODGA with metal ions have been illustrated. A

novel dimethyl dibutyl malonamide grafted polymer has been synthesized and

sorption behaviour of actinide ions on this grafted polymer has been described. The

basic as well as applied aspects of extraction of actinide ions with TODGA have

been explored. Various techniques employed for the separation of actinides / fission

products were solvent extraction, extraction chromatography and liquid membranes.

Synopsis

xvii

In conclusion, TODGA exhibited high basicity and high extraction capacity

for trivalent lanthanides / actinides as compared to commonly proposed extractants

such as CMPO and DMDBTDMA. TODGA forms third phase at very low

concentration of Nd, however, DHOA has been evaluated as a suitable phase

modifier. The possible application of TODGA for the separation of actinides /

lanthanides from radioactive waste solutions has been successfully demonstrated on

large scale in counter-current extraction mode using a mixer-settler system. The

extraction chromatographic studies involving TODGA as the stationary phase

demonstrated the possible use of the material for the concentration of radionuclides

from a large volume of dilute waste solutions. The sorption behaviour of uranium

and thorium on malonamide grafted polymer was found to follow the first order

Lagergren rate kinetics. The sorption of uranium on malonamide grafted polymer

exhibited the Langmuir adsorption isotherm. The Langmuir monolayer adsorption

phenomenon was also confirmed by the theoretical approach based on adsorption

kinetics. The transport behaviour of radionuclides by TODGA liquid membrane has

been described with the help of various diffusional parameters. Distilled water has

been evaluated as a suitable strippant for actinides / fission products. Stability of the

TODGA liquid membrane was found to be excellent when monitored over a period

of twenty days of continuous operation. The possible application of TODGA-based

liquid membrane for the separation of metal ions on large scale has been

demonstrated using hollow fibre membrane modules.

REFERENCES

1. Status and Trends of Spent Fuel reprocessing, IAEA TECDOC-1103, 1999. 2. L.H. Baestle. Burning of Actinides: A complementary waste management

option? IAEA Bulletin, 34(3) (1992), 32. 3. J.N. Mathur, M.S. Murali and K.L. Nash, Solv. Extr. Ion Exch., 19 (2001) 357.

4. V.K. Manchanda and P.N. Pathak, Sep. Purif. Technol., 35 (2004) 85. 5. L. Spjuth, J.O. Liljenzin, M.J. Hudson, M.G.B. Drew, B.P. Iveson and C. Madic,

Solv. Extr. Ion Exch., 18 (2000) 1. 6. Y. Sasaki, Y. Sugo, S. Suzuki and S. Tachimori, Solv. Extr. Ion Exch., 19 (2001)

91.

Synopsis

xviii

7. P.R. Danesi, E.P. Horwitz and P.G. Rickert, J. phys. Chem., 87 (1983) 4708. 8. L. Nunez, B.A. Buchholz and G.F. Vandergrift, Sep. Sci. Technol., 30 (1995)

1455.

9. J.L. Cortina and A. Warshawsky, developments in solid-liquid extraction by

solvent impregnated resins, In Ion exchange and solvent extraction, J.A.

Marinsky and Y. Marcus (Eds.), Marcel Dekker, NY (1975), Vol. 13, P. 195-293. 10. V. Camel, Spectrochim. Acta Part B, 58 (2003) 1177.

11. N. Masque, R.M. Marce and F.B. Trends, Anal. Chem., 17 (1998), 384. 12. C. Musikas, Inorg. Chim. Acta, 140 (1987) 197. 13. G. Spach, Ed., "Physical Chemistry of Transmembrane Ion Motions", Elsevier:

Amsterdam, 1983.

14. R.M. Izatt, J.D. Lamb and R.L. Bruening, Sep. Sci. Technol., 23 (1988) 1645. 15. N.M. Kocherginsky, Q. Yang and L. Seelam, Sep. Purif. Technol., 53 (2007) 171.

------------------------

Chapter I

General Introduction

Chapter I

1

2%7%

16%17%

19%

39%

Coal Gas Nuclear Hydro Oil Others

GENERAL INTRODUCTION

Our planet is witness to a constant increase in the population with a corresponding

increase in the needs of each individual. The demands for agricultural and industrial

output and essential services can only be met if the production of power (energy)

increases rapidly. While it is forecasted that the electrical power production in

industrialized countries will have to be doubled within the next 20years, the growth

rate of power generation will have to be much higher for developing countries like

India. At present, vast bulk of the global energy is supplied by coal, natural gas,

hydroelectric and, to a small extent, by oil and nuclear energy (Fig. 1.1). Due to

limited resources of the fossil fuels the overwhelming demand of global energy can

only be achieved by utilization of other possible resources. Nuclear energy has been

projected as an alternate source to meet the considerable energy requirement of the

world.

1.1. NUCLEAR ENERGY The nuclear power is characterized by the release of very large amount of energy

from a given amount of fuel generating relatively small amount of waste per unit

Fig. 1.1. World production of electricity in 2002 by various fuels. Source: OECD/IEA world energy outlook 2004

Chapter I

2

production of electrical energy. The basic nuclear reaction, viz. neutron induced

fission of fissile materials like 235U, results in the release of enormous amount of

energy. This fundamental nuclear reaction is utilized to obtain the controlled release

of energy in the nuclear power reactors. However, due to limited natural resources of

the fissile material (235U), the future nuclear energy program largely depends upon

the availability of the man made fissile materials like 233U and 239Pu. To sustain

nuclear power programme beyond the availability of naturally occurring 233U, it is

imperative to follow the closed fuel cycle option. The closed fuel cycle emphasizes

on the recycling of the spent fuel and has been already opted by several nations

including India. During reprocessing of the spent fuel in the closed fuel cycle, the

valuable plutonium and uranium are recovered by the hydrometallurgical process

leaving behind highly radioactive liquid waste solution, referred to as High Level

Waste (HLW). The HLW solution contains long-lived alpha emitting actinides such

as 241Am, 243Am, 245Cm and 237Np (referred to as minor actinides) apart from the

small amount of un-recovered plutonium and uranium as well as beta / gamma

emitting fission products and significant concentrations of structural materials and

process chemicals [1,2]. Since the half lives of minor actinides and some of the

fission products range from few hundred to millions of years, the HLW possesses

long term radiological risk to the environment [3]. The sustainability of the future

nuclear energy program, therefore, depends upon the safe management of radioactive

waste which shall never jeopardize the human health as well as the ecology. For

efficient radioactive waste management it is desirable to understand the source and

composition of radioactive waste generated at various stages of the nuclear fuel

cycle.

1.2. NUCLEAR FUEL CYCLE Nuclear fuel cycle comprises of front end and back end and comprises of various

stages like mineral exploration, mineral processing, purification of uranium /

thorium, fuel fabrication, reactor operation, spent fuel reprocessing, radioactive waste

management etc. (Fig. 1.2). The Front End includes stages from mining of the ore

to the reactor operation, and the Back End includes the removal of spent fuel from

the reactor and its subsequent reprocessing to recover valuables, and treatment and

disposal of high level waste.

Chapter I

3

Fig. 1.2. Nuclear Fuel Cycle

1.2.1. Waste from Front End of Fuel Cycle The waste generated at the uranium mine site comprises decay products of 238U / 233U

and exists in the form of radioactive dust. At the mill, dust is collected and fed back

into the process, while radon gas is diluted and dispersed into the atmosphere. The

wastes from the milling operation include the radioactive radium which is reverted

back to the mine and covered with rock and clay. The uranium oxide produced from

the mining and milling of the ore is accompanied by only a fraction of total

concentration of decay products as most of them are diverted to the tailings.

Similarly, the step of turning uranium oxide concentrate into a useable fuel does not

produce significant radioactive waste. It is when uranium is burnt in the reactor that

significant quantities of highly radioactive fission / activation products are produced

(Table 1.1). More than 99.9% of the radioactivity produced in the reactor is retained

in the fuel rods, while less than 0.1% is distributed in other systems of the reactor.

Chapter I

4

Table 1.1: Major contributors to the radioactivity in the spent fuel after a cooling period of 50 days

Nuclides Half life Nuclides Half life 3H 12.3 yrs 131I 8.05 days

85Kr 10.8 yrs 137Cs 30.0 yrs 89Sr 50.6 days 140Ba 12.8 days 90Sr 28.8 yrs 140La 40.2 days 90Y 64.4 hrs 141Ce 32.4 days 91Y 58.8 days 143Pr 13.6 days 95Zr 65 days 144Ce 285 days 95Nb 35 days 144Pr 17.3 min 103Ru 39.6 days 147Nb 11.1 days 106Ru 367 days 147Pm 2.62 yrs

129mTe 34 days

1.2.2. Waste from Back End of Fuel Cycle In the nuclear fuel cycle most of the radioactive waste is generated during

reprocessing of the spent fuel, i.e. at the back end of the fuel cycle. The fuel after

sufficient use in the reactor is referred as Spent Fuel. This irradiated spent fuel

contains long-lived alpha emitting transuranic elements (principally Np, Pu, Am and

Cm), which are formed in uranium fuelled reactors by neutron capture of 238U

followed by a sequence of beta emission and neutron capture reaction of the daughter

products. Apart from this, the spent fuel also contains large amount of fission

products which are generally beta/gamma emitters and constitute major dose in the

waste [2]. Although nearly 200 radionuclides are produced during irradiation of the

fuel, the great majorities of them are relatively short lived and decay to low level

within few decades. The major contributors to the fission product activity after a

cooling period of 50 days are listed in Table 1.1. The spent fuel is often allowed to

cool for few years to allow short lived radionuclides to decay. After cooling the spent

fuel for about one year, only 106Ru, 106Rh, 90Sr, 90Y, 144Ce, 144Pr, 134Cs, 137Cs and 147Pm contribute significantly to the activity [2]. During reprocessing of spent fuel

Chapter I

5

Fig. 1.3. Reprocessing of spent fuel

the irradiated fuel is dissolved in nitric acid solution, referred as dissolver solution,

and subsequently treated with tributyl phosphate (PUREX Process) to remove

valuable plutonium and uranium. A flow sheet for reprocessing of the spent fuel is

shown in Fig. 1.3. The aqueous raffinate remaining after the co-extraction of uranium

and plutonium from dissolver solution by PUREX process is concentrated into high

acidic liquid solution which is referred as High Level Waste (HLW). The HLW

solution thus contains minor actinides, fission products and left over uranium and

plutonium along with structural materials and process chemicals. One of the

challenges at the back end of the nuclear fuel cycle lies in the safe management of

HLW. Some of the radionuclides in HLW are very important and precious and hence

can be separated as wealth from the waste.

1.3. CLASSIFICATION OF RADIOACTIVE WASTE Radioactive wastes are classified as low level waste, intermediate level waste and

high level waste depending upon the level of radioactivity which varies from curies

per litre to microcuries per litre.

Chapter I

6

1.3.1. Low Level Waste When the total radioactivity of the waste is less than millicurie / litre, it is referred as

low level waste (LLW). It is generated as liquid from the decontamination of

equipments, radioactive laboratories, hospitals using radiopharmaceuticals as well as

from the nuclear fuel cycle. The level of radioactivity and half-lives of radioactive

isotopes present in LLW are relatively small. Storing the waste for a period of few

months allows most of the radioactive isotopes to decay, the point at which the

wastes can be disposed off safely. The LLW comprises about 90% of the total

volume of the radioactive wastes generated, but only < 1% radioactivity of all the

wastes. To reduce the volume of solid LLW, it is often incinerated and compressed

before disposal. Usually it is buried in shallow landfill sites.

1.3.2. Intermediate Level Waste When the radioactivity of the waste ranges from millicurie to curie / litre, the waste is

referred as intermediate level waste (ILW). The ILW contains higher amount of

radioactivity as compared to the LLW and, therefore, may require special shielding.

It typically comprises resins, chemical sludges, reactor components as well as

reprocessing equipments. The ILW comprises about 7% of the total volume of the

radioactive wastes, while it contains < 4% radioactivity of all the radioactive wastes.

1.3.3. High Level Waste When the radioactivity of the waste is greater than curie / liter, the radioactive waste

is referred as high level waste (HLW). The HLW is the waste emanating from the

reprocessing of spent fuel. While HLW comprises only about 3% of the total volume

of all the radioactive wastes, it contains more than 95% of the total radioactivity

generated in the nuclear fuel cycle. This waste includes uranium, plutonium and

other highly radioactive elements made up of fission products and alpha emitting

minor actinides. The challenge for the final disposal of HLW is largely due to the

radiotoxicity associated with the minor actinides which have half lives ranging from

few hundred to millions of years [4]. Efforts are being made by radiochemists /

separation chemists to meet the challenges of radioactive waste management by

developing efficient and environmentally benign processes for the separation of

Chapter I

7

101 102 103 104 105 10610-2

10-1

100

101

102

103

104

105

Parti

tioni

ng

Uranium Ore

No Partitioning

Radi

otox

icity

(Rel

ativ

e)

Time (Years)

various radionuclides from HLW solution. This would minimize the volumes of

radioactive wastes and costs of their final disposal.

1.4. IMPACT OF RADIONUCLIDES ON ENVIRONMENT The long-lived radionuclides present in the raffinate of PUREX process after

reprocessing of the spent fuels are of great environmental concern. The radioactive

waste, whether natural or artificial, is a potential source of radiation exposure to the

human being through different pathways. The raffinate after PUREX process

generally contains un-extracted U, Pu and bulk of minor actinides such as Am, Np,

Cm and host of fission products like Tc, Pd, Zr, Cs, Sr and lanthanides as well as

activation products. At present the most accepted conceptual approach for the

management of HLW is to vitrify it in the glass matrix followed by disposal in deep

geological repositories [5,6]. Since the half lives of minor actinides concerned range

between a few hundred to millions of years, the surveillance of high active waste for

such a long period is debatable from economical as well as environmental safety

considerations. On the other hand, the vitrified mass of HLW will have to withstand

the heat and radiation damages caused by the decay of beta/gamma emitting fission

products such as 137Cs and 90Sr for about 100yrs. Therefore, it may create the

possible risk for the migration of long lived alpha emitting minor actinides from

Fig. 1.4. Partitioning of minor actinides- Impact on waste management

Chapter I

8

repository to the environment. The recommended activity level of 4000Bq per gram

in terms of alpha activity is considered benign enough to be treated as LLW. As

represented in Fig. 1.4, if actinides are not removed from the spent fuel, it will

require millions of years to reduce its radiotoxicity to this level. However, if one can

remove U, Pu and minor actinides from the waste its radiotoxicity could reach an

acceptable level after few hundreds of years. Therefore, strategy of P&T (Partitioning of long-lived radionuclides followed by Transmutation) is being

considered by several countries around the world [7,8]. The P&T process envisages

the complete removal of minor actinides from radioactive waste and their subsequent

burning in the reactors / accelerators as mixed oxide fuel. This process will lead to

generation of extra energy and at the same time would alleviate the need for long

term surveillance of geological repositories. After partitioning of the actinides along

with the long lived fission products, the residual waste can be vitrified and buried in

subsurface repositories at a much reduced risk and cost.

1.5. CHEMISTRY OF ACTINIDES The work carried out in this thesis pertains to the separation chemistry of actinides

and fission products from radioactive waste solutions. The actinides include uranium,

neptunium, plutonium, americium and curium. It is quite essential to understand the

chemistry of actinides before their partitioning. A brief survey of the chemistry of

actinide elements is, therefore, considered relevant.

1.5.1. History The existence of rare earth like series in the seventh row of periodic table, which was

suggested as early as 1926, gained wider acceptance with the discovery and study of

transuranium elements [9]. In 1945, Seaborg proposed that actinium and

transactinium elements form such a series in which the 5f electron shell is being

filled in a manner analogous to the filling of 4f shell in lanthanides [10]. Except for

uranium and thorium, which are well known actinide elements discovered in 1789

and 1828, respectively, all the other elements were discovered in twentieth century.

Among actinide elements uranium and thorium have isotopes with half-lives

exceeding the estimated life of this planet and hence occur in nature. Actinium and

protactinium owe their existence to the decay of long lived isotopes of uranium,

Chapter I

9

thorium and their daughter products. The rest of the elements in this series are

essentially man made with some evidence for the trace occurrence of neptunium

and plutonium in the nature formed by nuclear reactions involving uranium [11,12].

Among man made elements plutonium and, to a lesser extent, neptunium, americium

and curium are produced in the nuclear power reactors and are recovered from the

spent nuclear fuels. The elements beyond curium are generally produced through

heavy ion reactions of transplutonium elements in accelerators. With increasing

atomic number of actinides, the nuclei becomes rapidly less stable and only

einsteinium has an isotope with a half-life long enough to offer any possibility for

conventional chemical studies.

1.5.2. Electronic Configuration The fourteen 5f electrons enter the actinide elements beginning formally with Th

(Z=90) and ending with Lr (Z=103). These fourteen elements following Ac are

placed in the 7th row of the periodic table separately analogous to lanthanides.

Intensive chemical studies have revealed many similarities between the lanthanides

and actinides. The ground state electronic configuration of lanthanides and actinides

is shown in Table 1.2. Though there is over all similarity between the two groups of

elements, some important differences also exist mainly because the 5f and 6d shells

are of similar energy in actinides and 5f electrons are not so well shielded as 4f

electrons in lanthanides [13]. The lighter actinides (Ac to Np) show greater tendency

to retain 6d electrons due to smaller energy differences between 6d and 5f orbitals

relative to that between 5d and 4f orbitals of lanthanides. In case of transition series

the relative energy of orbitals undergoing the filling process become lower as the

successive electrons are added. For actinides too the 5f orbitals of plutonium and

subsequent elements are of lower energy than 6d orbitals and, therefore, the

subsequent electrons are filled in 5f orbitals with no electrons in 6d orbitals.

1.5.3. Solution Chemistry of Actinides As the processes of separation and purification of actinides on large scale are

essentially based on hydrometallurgical techniques, the study of solution chemistry

of actinides has received considerable attention. The actinide elements exist in

multiple oxidation states and most of their separation processes are based on the

Chapter I

10

Table 1.2: Electronic configuration of lanthanide and actinide elements

Lanthanides Actinides

Elements Atomic numbers

Electronic configurations

Elements Atomic numbers

Electronic configurations

La 57 5d1 6s2 Ac 89 6d1 7s2

Ce 58 4f 1 5d1 6s2 Th 90 6d2 7s2

Pr 59 4f 3 6s2 Pa 91 5 f 2 6d1 7s2

Nd 60 4f 4 6s2 U 92 5f 3 6d1 7s2

Pm 61 4f 5 6s2 Np 93 5f 4 6d1 7s2

Sm 62 4f 6 6s2 Pu 94 5f 6 7s2

Eu 63 4f 7 6s2 Am 95 5f 7 7s2

Gd 64 4f 7 5d1 6s2 Cm 96 5f 7 6d1 7s2

Tb 65 4f 9 6s2 Bk 97 5f 9 7s2

Dy 66 4f 10 6s2 Cf 98 5f 10 7s2

Ho 67 4f 11 6s2 Es 99 5f 11 7s2

Er 68 4f 12 6s2 Fm 100 5f 12 7s2

Tm 69 4f 13 6s2 Md 101 5f 13 7s2

Yb 70 4f 14 6s2 No 102 5f 14 7s2

Lu 71 4f 14 5d1 6s2 Lr 103 5f 14 6d1 7s2

effective exploitation of these properties. It is, therefore, desirable to understand the

various oxidation states of actinides in solution.

1.5.3.1. Oxidation States

The trivalent oxidation state is the most stable for all lanthanides. However, this is

not so at least in the case of earlier members of actinide series. The 5f electrons of

actinides are subjected to a lesser attraction from the nuclear charge than the

corresponding 4f electrons of lanthanides. The greater stability of tetra positive ions

of early actinides is attributed to the smaller values of fourth ionization potential for

5f electrons compared to 4f electrons of lanthanides, an effect which has been

observed experimentally in the case of Th and Ce [14]. Thus, thorium exists in

aqueous phase only as Th(IV) while the oxidation state 3+ becomes dominant only

Chapter I

11

Table 1.3: Oxidation states* of actinide elements

89 90 91 92 93 94 95 96 97 98 99 100 101 102 103

Ac Th Pa U Np Pu Am Cm Bk Cf Es Fm Md No Lr

(2) (2) 2 2

3 (3) (3) 3 3 3 3 3 3 3 3 3 3 3 3

4 4 4 4 4 4 4 4

5 5 5 5 5

6 6 6 6

7 7

* Those underlined are the most stable oxidation states in aqueous solution; those in parentheses refer to oxidation states which are not known in solutions. for transplutonium elements. The actinides existing in different oxidation states are

shown in Table 1.3, where the most stable oxidation states are under lined [13]. All

the oxidation states are well known except 7+ states for Np and Pu which exist in

alkaline medium[15]. Penta and hexavalent actinide ions exist in acid solution as

oxygenated cations, viz. MO2+ and MO22+.

Fig. 1.5. Redox potential of actinide ions in 1M HClO4 (Volts)

Chapter I

12

The redox potential diagrams of early actinides such as Th, U, Np and Pu at

25C in 1M HClO4 are shown in Fig. 1.5 [16,17]. It has been found that the M3+/M4+

and MO2+/MO22+ couples are reversible and fast as they involve the transfer of only

single electron. On the other hand, the other couples are irreversible and achieve

equilibrium slowly as they involve the formation or rupture of metal oxygen bonds.

1.5.3.2. Disproportionation Reactions

Disproportionation reaction is referred to as self oxidation reduction reaction. For

disproportionation reaction to occur an element must have at least three oxidation

states and these ions must be able to co-exist in solutions, which depend on the

closeness of the electrode potentials of redox couples involved. In case of Pu these

values are so close that the four oxidation states, viz. III, IV, V and VI are in

equilibrium with each other. The disproportionation reactions of U, Pu, Np and Am

have been well studied [13] and their equilibrium constant (logK) values are given in

Table 1.4. In general, disproportionation reactions of MO2+ (M=U, Pu or Np) ions

can be represented as follows,

2MO2+ + 4H+ M4+ + MO22+ + H2O (1.1)

Table 1.4: Disproportionation reactions of actinides in aqueous solutions

Element Oxidation Numbers Reaction logK (25C)

U V = IV + VI 2UO2+ + 4H+ U4+ + UO22+ + 2H2O 9.30

Np V = IV + VI 2NpO2+ + 4H+ Np4+ + NpO22+ + 2H2O -6.72

Pu V = IV + VI 2PuO2+ + 4H+ Pu4+ + PuO22+ + 2H2O 4.29

V = III + VI 3PuO2+ + 4H+ Pu3+ + 2PuO22+ + 2H2O 5.40

IV + V = III + VI Pu4+ +PuO2+ Pu3+ + PuO22+ 1.11

IV = III + VI 3Pu4+ + 2H2O 2Pu3+ + PuO22+ +4H+ -2.08

Am IV + V = III + VI Am4+ +AmO2+ Am3+ + AmO22+ 12.5

IV = III + VI 3Am4+ + 2H2O 2Am3+ + AmO22+ +4H+ 32.5

IV = III + V 2Am4+ + 2H2O Am3+ + AmO2+ +4H+ 19.5

Chapter I

13

It is clearly demonstrated from the equilibrium reaction (1.1) that the presence of

hydrogen ion and complexing ions like F- and SO42-, which complex strongly with

M4+ and MO22+ ions, have pronounced effect on disproportionation reactions.

1.5.3.3. Hydrolysis and Polymerization

In view of their large ionic potential, the actinide ions in various oxidation states

exist strongly as hydrated ions in the absence of complexing ions. The actinide ions

in divalent to tetravalent oxidation states are present as M2+, M3+ and M4+,

respectively. The penta and hexavalent oxidation states are prone to more hydrolysis

as compared to lower oxidation states. These oxidation states exist as partially

hydrolyzed actinyl ions, viz. MO2+ and MO22+ and can get further hydrolyzed under

high pH condition. The oxygen atoms of these ions are not basic in nature and thus

do not co-ordinate with protons. The degree of hydrolysis for actinide ions decreases

in the order: M4+ > MO22+ > M3+ > MO2+ which is similar to their complex formation

properties [18]. In general the hydrolysis of the actinide ions can be represented as

follows,

Mn+ + xH2O M(H2O)xn+ M(OH)x(n-x)+ + xH+ (1.2)

The hydrolysis behaviour of Th(IV) is quite different from those of other tetravalent

actinide ions [19]. For U(IV) and Pu(IV) the metal ion hydrolyses first in a simple

monomeric reaction (Eq. 1.2) followed by a slow irreversible polymerization of

hydrolyzed products. For Th(IV), however, various polymeric species exist even in

very dilute solutions. Whereas the polymer formation of Pu(IV) is irreversible, that

of Th(IV) is reversible. The hydrolysis of some of the trivalent actinides such as

Am(III), Cm(III) and Cf(III) is well studied which revealed the higher hydrolysis

constant values for trivalent actinides as compared to their lanthanides analogues

[13].

Though the polynuclear species of all actinide ions are of great interest, the

polymers of Pu(IV) have attracted particular attention because of practical

considerations. Pu(IV) polymers with varying molecular weights ranging from a few

thousand to as high as 1010 have been observed [20]. In dilute HNO3 or HCl

solutions, Pu(IV) polymer exists as a bright green colour with a characteristic

spectrum different from that of monomeric Pu(IV) in these solutions. The rate of

Chapter I

14

polymerization depends on acidity, temperature, Pu(IV) concentration as well as the

nature of ions present in the solution [21,22]. Polymerization rate for Pu(IV) is higher

when the ratio of acid to Pu(IV) concentration is low. Thus, Pu(IV) polymerization

can occur even at higher acidities if Pu(IV) concentration is raised. Depolymerization

of Pu(IV) is best accomplished by heating the Pu solution in 610M HNO3. Strong

complexing agents such as fluoride and sulphate ions as well as oxidizing agents

such as permanganate and dichromate promote depolymerization of plutonium.

1.5.3.4. Complexation of Actinides

The actinide ions in the aqueous solutions exhibit strong tendency to form

complexes. This property of actinides is widely exploited in devising methods for

their separation and purification. One of the most important factors that determines

the strength of the complex formed is the ionic potential (or charge density) of the

metal ions, which is the ratio of ionic charge to ionic radius. Higher the ionic

potential greater the electrostatic attraction between cations and anions and hence

stronger is the complex formed. The complexing strength of actinide ions in different

oxidation states follows the order: M4+ > MO22+ > M3+ > MO2+. Similarly, for the

given metal ions of same oxidation state, the complexing ability increases with the

atomic number due to increase in the ionic potential as a result of actinide contraction

[13]. However, the above generalized statement may be valid when complexation is

primarily ionic in nature. There are large number of instances where hybridization

involving 5f orbitals, steric effects and hydration of metal ions affect the tendency of

complexation. For anions the tendency to form complex with the given actinide ion

generally vary in the same manner as their abilities to bind with hydrogen ion [23].

For monovalent ligands the complexing tendency decreases in the order: F- >

CH3COO- > SCN- > NO3- > Cl- > Br- > I- > ClO4-. The divalent anions usually from

stronger complexes than the monovalent anions and their complexing ability

decreases in the order: CO32- > SO32- > C2O42- > SO42-. The complexing ability of

some of the organic ligands with Th(IV) varies as: EDTA > Citrate > Oxalate >

HIBA > Lactate > Acetate.

While discussing the stability of complexes between metal ions and ligands,

Pearson [24] proposed a scheme based on the concept of hard and soft acids and

bases. Those metal ions are called hard which have a small radius and high charge

Chapter I

15

and do not possess valence shell electrons that are easily distorted. The soft metal

ions have the opposite characteristics. When similar classification is applied to the

ligands it is observed that the hard metal ions form stronger complexes with hard

ligands and soft metal ions with soft ligands. Actinide ions behave as hard acids

and interact strongly with hard bases such as O or F rather than soft ligands like

N, S or P donors. However, as compared to lanthanides they show marked

preference for the soft donors which is commonly referred as covalent character due

to the f-orbital participation. The complex formation reactions involving hard acids

and bases are endothermic whereas the reverse is true for soft ions. This is because

the complex formation between hard metal ions and hard ligands require the breaking

of strong bonds between these metal ions and water molecules in the primary

hydration sphere which require large energy. The process of removal of water

molecules, however, results in large increase in entropy which contributes to the

driving force of these reactions [13]. When the primary hydration shell is broken

during complex formation, the complex formed is referred as inner sphere

complex. In contrast outer sphere complexes do not require breaking of the

primary hydration shell. The actinide ions interact with soft bases in organic solvents

of low solvating power, but not in aqueous solutions where the soft bases would have

to replace inner sphere water molecule which is a hard base. Thus, depending upon

the nature of ligand and medium actinide cations form inner or outer sphere

complexes.

1.5.3.5. Absorption Spectra

Similar to transition metal ions, the actinide ions display a rich variety of colours in

their aqueous solutions. The absorption spectra of actinides arise due to the electronic

transitions and absorption bands appear mainly from three types of transitions, viz. i)

f-f- transition, ii) f-d transition, and iii) charge transfer bands [13]. In f-f transitions,

the electronic transition occurs between the two 5f-5f orbitals of different angular

momentum. As the transitions occur between the orbitals of the same sub-shell they

are generally Laporate forbidden. The probabilities of transitions are, therefore, low

and the absorption bands are consequently low in intensity. However, the bands are

sharp because the transitions take place in the inner shell and are, therefore, not

affected much by the surrounding environment. The energy differences between the

Chapter I

16

various energy levels are of such an order of magnitude that the bands due to 5f-5f

transitions appear in UV, visible and near IR regions. The molar absorption

coefficient is in the range of 1050 M-1cm-1. On the other hand, in case of f-d

transitions the absorption bands are broad as these transitions are influenced by the

surrounding environment. As transitions take place between the orbitals of different

azimuthal quantum number they are Laporate allowed and, therefore, these bands are

relatively more intense. The molar absorption coefficient is of the order of ~10000

M-1cm-1. These bands appear invariably in the UV region due to large energy

differences between the d and f orbitals. In case of charge transfer transitions, the

absorption bands occur due to the transition between 5f orbitals of actinide ions and

ligand orbitals. Therefore, the nature of ligand plays an important role. These

transitions are significantly affected by the surrounding environment. As a

consequence, the charge transfer bands are broad. The absorption bands appear in the

UV region and are generally less intense than those resulting from f-d transitions.

The absorption spectra of actinide ions have been widely used in the

analytical chemistry. The absorption spectra of actinide ions in different oxidation

states differ widely, which have been successfully exploited for the quantitative

analysis of their mixtures present in different oxidation states. The absorption bands

of actinide ions have also been used for studying the redox reactions. Though the

transitions in actinide ions take place in an inner shell resulting in sharp bands,

complexing of metal ions can strongly affect the position as well as the intensities of

the individual absorption bands. Therefore, change in absorption spectra due to the

presence of ligands have often been used to establish complex formation, and in

some cases, even for the calculation of their stability constants. The complexes of

some of the actinides formed with many organic and inorganic ligands have very

high absorption in visible region. This property has been fruitfully exploited to

develop sensitive analytical methods for the detection and estimation of actinide ions.

1.6. SEPARATION OF METAL IONS The scientific principles that govern the separation of metal ions from solutions are

chemical reaction equilibrium kinetics, fluid mechanics and mass transfer from one

phase to another. The theory of separation utilizes these principles in different

processes including solvent extraction, extraction chromatography as well as in

Chapter I

17

membrane processes. Amongst these techniques, solvent extraction is the most

versatile technique and is extensively used for separation, preparation, purification,

enrichment and analysis on micro scale to large industrial processes.

Solvent or liquid-liquid extraction is based on the principle that a solute can

distribute itself in a certain ratio between the two immiscible solvents, one of which

is usually water and the other is an organic solvent. In certain cases the solute can be

more or less completely transferred into the organic phase. The liquidliquid

distribution systems can be thermodynamically explained with the help of phase rule

[25]. Phase rule is usually stated as,

P + V = C + 2 (1.3)

where P, V and C denote the number of phases, variances and components,

respectively. In general, a binary liquid-liquid distribution system has two phases (P

=2) and contains three or more components (two solvents and one or more solutes).

When a system contains only one solute (C = 3), according to the phase rule the

variance is three, which means by keeping any two variables constant the system can

be defined by the third variable. In other words, at fixed temperature and pressure,

the concentration of solute in the organic phase is dependent on the concentration of

solute in the aqueous phase. Thus, when molecular species of the solute is same in

the two phases, its concentration in one phase is related to that in the other phase (the

distribution law). Consider following equilibrium reaction,

M(aq.) M(org.) (1.4)

where the subscripts (aq.) and (org.) represent aqueous and organic phases,

respectively. According to the distribution law, the distribution coefficient (kd) is

represented as,

[Mn](org.) kd = ------------------ (1.5) [Mn](aq.)

However, it has been observed that, in most cases, the molecular species of metal

ions are not the same in both the phases. Therefore, the term distribution ratio (DM)

is used in the solvent extraction which is defined as the ratio of the total

concentration of metal ions (in all forms) in the organic phase to that of in the

aqueous phase.

Chapter I

18

The solubility of charged metal ions in the organic solvents are very less as

they tend to remain in the aqueous phase due to ion-dipole interaction. For the

extraction of metal ions in the organic phase, the charge on the metal ions must be

neutralized so as to enhance the solubility in non-polar organic solvents. Therefore, a

suitable extractant (ligand) molecule is generally added in the organic solvent which

upon complexation with metal ions forms neutral hydrophobic species which is then

extracted in the organic phase. In such cases, the extraction of metal ions may follow

one of the following extraction mechanisms. (i) Solvation: The extraction of metal

ions by neutral ligands are followed by solvation mechanism. The extraction process

proceeds via replacement of water molecules from the co-ordination sphere of metal

ions by basic donor atoms such as O or N of the ligand molecules. The well

known example is the extraction of U(VI) by tri-n-butyl phosphate (TBP) from nitric

acid medium [26]. (ii) Chelation: The extraction of metal ions proceeds via the

formation of metal chelates with chelating ligands. The example of this type is the

extraction of Pu(IV) by thenoyltrifluoroacetone (HTTA) in benzene [27]. (iii) Ion

pair extraction: This type of extraction proceeds with the formation of neutral ion-

pair species between the metal ions and ionic organic ligands. Acidic ligands such as

sulphonic acids, carboxylic acids and organophosphoric acids provide anions by

liberating protons which then complexed with the metal cation to form ion-pair. On

the other hand, basic ligands provide cations which complex with aqueous anion

metal complex to form ion-pair. The best examples of basic extractants are

quaternary ammonium salts. (iv) Synergistic extraction: Synergism refers to the

phenomenon where the extraction of metal ions in the presence of two or more

extractants is more than that expected from the sum of extraction employing

individual extractants. Well known example of synergistic extraction is the extraction

of Pu(IV) from nitric acid medium by a mixture of HTTA and tri-n-octyl phosphine

oxide (TOPO) in benzene [28].

1.7. CRITERIA FOR SELECTION OF EXTRACTANTS A number of factors are taken into consideration while selecting or designing a

particular extractant for the separation of metal ions for industrial applications [29].

Some of the important considerations are listed as follows,

i) High solubility in paraffinic solvents (non-polar solvents),

Chapter I

19

ii) Low solubility in the aqueous phase,

iii) Non-volatility, non-toxicity and non-inflammability,

iv) High complexation ability with the metal ions of interest,

v) High solubility of the metal-ligand complex in the organic phase, i.e. high

metal loading capacity in the organic phase,

vi) Ease of stripping of metal ions from the organic phase,

vii) Reasonably high selectivity for the metal ion of interest over the other metal

ions present in the aqueous solution,

viii) Optimum viscosity for ease of flow and optimum inter facial tension (IFT) to

enable a faster rate of phase disengagement,

ix) Ease of regeneration of the extractant for recycling,

x) High resistance to radiolytic and chemical degradation during operation, and

xi) Ease of synthesis / availability at a reasonable cost.

1.8. REPROCESSING OF SPENT FUEL The fuel after use in the reactor is referred to as spent fuel. The spent fuel contains

man made fissile materials such as 239Pu along with minor actinides and fission

products. Reprocessing of the spent fuel is important for the recovery of valuable

fissile materials to sustain the future nuclear energy programme. During reprocessing

of the spent fuel the valuable uranium and plutonium are recovered in the

hydrometallurgical process leaving behind highly radioactive liquid waste solution

(HLW). A brief mention about the reprocessing of the spent fuel by PUREX process

is presented here.

1.8.1. PUREX Process The Plutonium Uranium Reduction Extraction (PUREX) process is employed for

reprocessing of the spent nuclear fuel throughout the world [30]. It involves

contacting a nitric acid solution of dissolved irradiated fuel with an organic solution

of tri-n-butyl phosphate (TBP) in a hydrocarbon diluent such as odourless kerosene

or n-dodecane. Typically, the TBP concentration is about 30% though the

concentration may be varied to effect a specific separation. The PUREX process is

based on the fact that TBP selectively extracts hexavalent uranium and tetravalent

plutonium over other actinides and fission products from moderately concentrated

Chapter I

20

(~3M) nitric acid solutions. By adjusting the valency of plutonium from tetravalent to

trivalent it may be partitioned from the organic phase to the aqueous solution, thus

providing an effective mean of separating plutonium from uranium. If the TBP

solution of extracted actinides is subsequently contacted with dilute (~0.1M) nitric

acid, the U(VI) may be easily back extracted (stripped) into the aqueous phase.

Though the PUREX process applied for reprocessing of the spent fuel

removes all uranium and plutonium, it rejects trivalent (Am and Cm) and pentavalent

(Np) actinides along with the fission products towards the aqueous raffinate. The

challenges for the disposal of aqueous raffinate generated during the PUREX process

(HLW) is largely due to the radiotoxicity associated with the transuranic actinides.

Thus, there is a need for the subsequent treatment of the aqueous raffinate to remove

all transuranic actinides before its disposal. Though PUREX process does not remove

all the actinides to the level necessary for their disposal (the process preferentially

recovers major actinides (U/Pu) present at ton/Kg scale leaving behind minor

actinides (Am/Cm/Np) present at mg/gm scale), it could provide a suitable clean feed

stream for the subsequent more efficient process of actinide partitioning [6].

1.9. ACTINIDE PARTITIONING The selective extraction of trivalent actinides, namely Am(III) and Cm(III) present in

the HLW resulting from the reprocessing of the spent fuel is influenced by the

presence of trivalent lanthanides. The trivalent lanthanides have almost similar

chemical properties to those of trivalent actinides and have several times higher

concentration than the later, which represent about 1/3 of the total mass of the fission

products. So owing to the complexity of the selective removal of trivalent actinides,

the separation process can be split into two steps. The first step consists of co-

extraction of An(III) and Ln(III) aiming to eliminate all the alpha activities and 1/3 of

the fission products. The second step consists of the group separation of Ln(III) and

An(III) by several processes including Selective Actinide Extraction (SANEX)

process. During the last two decades, concerted research conducted around the world

has identified a number of promising extractants for actinide partitioning. The

performance and status of some of these extraction processes are briefed here.

Chapter I

21

Fig. 1.6. Structural formulae of some of the proposed extractants for actinide partitioning

1.9.1. TRUEX Process The Trans Uranium Extraction (TRUEX) is a solvent extraction process designed to

separate transuranic elements from various types of high level waste solutions. The

key ingredient in this process is a phosphine oxide based extractant, viz.

octyl(phenyl)-N,N-diisobutyl carbamoyl methyl phosphine oxide (CMPO, Fig.

1.6(a)). Among several derivatives of phosphine oxide extractants, CMPO was found

(a) CMPO

(d) DMDBTDMA

(b) TRPO (R: n-octyl and n-hexyl)

(e) Tetra alkyl diglycolamide

(c) DIDPA

N

C

CH2

O

CH2

CH2

P

C8H17

O

CH

CH

H3C

H3C

H3C

H3C

P

OR1

R3R2

P

O

OOH

O

H2C

CH2

H2C

CH2

H2C

CH2

CH

CH3

H3C

CH2H2C

CH2CH2

H2C

CH2

H2C

CH

CH3

H3C

CH2

CHC C

N N

O O

C4H9

CH3

H3C

C4H9 C14H29

C

CH2

O

N

O R3

R4

CH2

CN

R1

R2

O

Chapter I

22

to possess the best combination of properties for actinide partitioning in a PUREX

compatible diluent system [31]. The TRUEX extractant is usually 0.2M CMPO +

1.2M TBP (used as a phase modifier) in paraffinic hydrocarbon like n-dodecane [32].

In TRUEX solvent, TBP suppresses third phase formation, contributes to better acid

dependencies for DAm, improves phase compatibility, and reduces hydrolytic and

radiolytic degradation of CMPO [33]. High distribution ratio of tri-, tetra- and

hexavalent actinides from solutions of moderate acid concentration and good

selectivity over fission products is the key feature of this extractant. Lanthanides such

as Eu, Ce and Pr behave similar to the trivalent actinides, viz. Am(III). Other fission

products, except Zr, show relatively small distribution values. Zirconium is also

extractable with TRUEX solvent; however, its extraction may be suppressed by the

addition of oxalic acid. From the process perspective, the insensitivity of distribution

values of actinides between 1M and 6M HNO3 is important as it allows efficient

extraction of these ions from waste with little or no adjustment of feed acidity.

Due to high extraction of tetra- and hexavalent actinides such as Pu(IV) and

U(VI) by CMPO in a wide range of acidity the stripping of these metal ions with

dilute nitric acid is difficult. Therefore, more aggressive stripping, for example with

powerful diphosphonate actinides extractants, is required. Generally, 1-

hydroxyethylene-1,1-diphosphonic acid (HEDPA) is used for stripping of Am, Pu

and U from loaded organic phase. The oxidation state specific stripping of actinide

ions from loaded TRUEX solvent can be achieved in three steps: 0.04M HNO3 to

remove trivalent actinides, dilute oxalic acid for selective stripping of tetravalent

actinides, and finally 0.25M Na2CO3 for uranium recovery. A mixture of formic acid,

hydrazine hydrate and citric acid has shown promise for efficient stripping of Am

and Pu from TRUEX solvent loaded with HLW in both batch as well as counter

current modes [34,35].

Though CMPO shows high extraction efficiency and is a promising reagent

for the separation of actinides, TRUEX process exhibits certain limitations. Stripping

of trivalent actinides is cumbersome and requires several stages of contact with

0.04M HNO3. Degradation products of CMPO can also inhibit the stripping of Pu

and U. The presence of acidic extractants as degradation products increases the DAm

values under stripping conditions. Such impurities must be removed from the used

Chapter I

23

TRUEX solvent prior to their recycling. More stringent stripping condition of metal

ions from the loaded organic phase is the major draw back of the TRUEX process.

1.9.2. TRPO Process Trialkyl Phosphine Oxide (TRPO) process utilizes a mixture of four alkyl phosphine

oxides (Fig. 1.6(b)) as the extractant. The TRPO solvent has been tested for the

extraction of actinides, lanthanides and other fission products from HNO3 and HLW

solutions [36,37]. It was observed that >99% of U(VI), Np(IV), Np(VI) and Pu(IV)

were extracted from 0.21M HNO3 through a single extraction with 30% (v/v) TRPO

in kerosene [38]. Also >95% of Pu(III), Am(III) and Ln(III) could be extracted, while

fission products such as Cs, Sr, Ru were not extracted. Trivalent lanthanides and

actinides are generally stripped with 5M HNO3. On the other hand, tetravalent (Np

and Pu) and hexavalent (U) actinides are stripped with 0.5M oxalic acid and 5%

Na2CO3, respectively. Though TRPO, with its relatively low cost and its high extraction efficiency,

is a promising extractant for actinide partitioning the process, however, it has certain

limitations. The TRPO process works only at relatively low acidity (0.1-1M HNO3)

and, therefore, the HLW solution (HLW is generally at ~3M HNO3) has to be diluted

several times to adjust the feed acidity. Poor stripping of actinide ions is also a

disadvantage of the TRPO process.

1.9.3. DIDPA Process The extraction behaviour of actinides and other fission products with di-isodecyl

phosphoric acid (DIDPA, Fig. 1.6(c)) has been studied by Morita et al., at Japan

Atomic Energy Research Institute (JAERI). It has been shown that DIDPA can

simultaneously extract Am(III), Cm(III), U(VI), Pu(IV) and even Np(V) from a

solution of low acidity such as 0.5M HNO3 [39,40]. The trivalent cations can be

separated from their tetravalent counterparts by appropriate back-extraction

procedures. The back extraction of trivalent actinides and lanthanides can be

achieved by 4M HNO3. On the other hand, tetravalent Np and Pu and hexavalent

uranium can be stripped by 0.8M oxalic acid and 1.5M Na2CO3 solution,

respectively. For the partitioning of transuranic elements a mixture of 0.5M DIDPA

+ 0.1M TBP in dodecane has been proposed.

Chapter I

24

The major drawback of DIDPA process is the re-adjustment of the acidity of

HLW to about 0.5M HNO3 prior to the processing. In this process, the reduction of

acidity and denitration is accomplished using formic acid. At such a low acidity,

molybdenum and zirconium form precipitates which carries about 90% of plutonium.

1.9.4. DIAMEX Process Diamide extraction (DIAMEX) process was developed in France for the extraction of

transuranic elements from the HLW solutions. One of the major drawbacks of using

organophosphorus extractants is the solid residue that results upon their incineration

at the end of their useful life. French researchers utilized the CHON (carbon,

hydrogen, oxygen and nitrogen) principle for designing of the extractants, which can

be completely incinerated into gaseous products, thereby minimizing the generation

of solid secondary wastes at the end of the process.

Among the numerous diamides synthesized and tested for the extraction of

actinides, N,N-dimethyl-N,N-dibutyl tetradecyl malonamide (DMDBTDMA, Fig.

1.6(d)) has shown the greatest promise [41-46]. In France, this reagent is extensively

evaluated for actinide partitioning from HLW solution. DMDBTDMA dissolved in

dodecane does not give any third phase when contacted with 3-4M HNO3 and hence

discourage the use of any phase modifier. Generally, 1M DMDBTDMA has been

proposed for actinide partitioning which gives DAm value of ~10 at 3M HNO3 [45].

Zirconium(IV) is strongly extracted by DMDBTDMA, however, its extraction can be

suppressed to an acceptable level by complexing it with oxalic acid. Extraction of

molybdenum can be suppressed by complexation with hydrogen peroxide. Iron,

which is almost always present in HLW from corrosion of the process equipments,

also has high affinity for DMDBTDMA. However, the extraction kinetics for Fe(III)

is slow and it may be separated from actinides and lanthanides by the judicial choice

of contact time for their extraction [45].

Recently, a new diamide, viz. N,N-dimethyl-N,N-dioctyl-2-(2-hexylethoxy)

malonamide (DMDOHEMA) has been reported as a substitute of DMDBTDMA for

DIAMEX solvent [47]. Amongst several extractants described for actinide

partitioning, diamides have been found to be particularly promising in view of their

improved back extraction properties for Am(III)/Cm(III), their complete

incinerability, and the innocuous nature of their radiolytic and hydrolytic products

Chapter I

25

(mainly carboxylic acids and amines) that can be easily washed out. However, the

major draw back of DMDBTDMA is that it shows only moderate extraction of

trivalent actinides (Am and Cm) from HLW at acidity 3M HNO3 [46]. Therefore, it

necessitates the structural modification of diamides so as to enhance the extraction

efficiency of trivalent actinides in particular.

1.10. DIGYLYCOLAMIDES: A Class of Promising Extractants for

Actinide Partitioning The performance of some of the extraction processes developed for actinide

partitioning is briefly discussed in the earlier section. However, each of the described

processes has certain limitations. The main drawbacks of the TRUEX process are; (a)

the poor back extraction of Am(III) and Cm(III) at reduced acidity, and (b)

interference due to solvent degradation products. On the other hand, the TRPO

process works only at relatively low acidity (0.1-1M HNO3) and, therefore,