-
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
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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
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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
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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
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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
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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
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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
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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.
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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)
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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
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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
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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
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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,
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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
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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.
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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
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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,