RECOVERY OF PURE PLUTONIUM
BY EXTRACTION WITH TRILAURYLAMINE AND "DIRECT PRECIPITATION"RECOVERY OF PURE PLUTONIUM
BY EXTRACTION WITH TRILAURYLAMINE AND "DIRECT PRECIPITATION"PROEFSCHRIFT
TER VERKRIJGING VAN DE GRAAD VAN DOCTOR IN DE TECHNISCHE WETENSCHAPPEN AAN DE TECHNISCHE HOGESCHOOL DELFT, OP GEZAG VAN DE RECTOR MAGNIFICUS DR. IR. C. J. D. M. VERHAGEN, HOOGLERAAR IN DE AFDELING DER TECHNISCHE NATUURKUNDE, VOOR EEN COMMISSIE UIT DE SENAAT TE VERDEDIGEN OP WOENSDAG
24 JANUARI 1968 TE 16.00 UUR
DOOR
JACOBUS NICOLAAS CORNELIS VAN GEEL
SCHEIKUNDIG INGENIEURGEBOREN TE EINDHOVEN
DRUKKERIJ UITGEVERIJ H. GIANOTTEN N.V. - TILBURG
DIT PROEFSCHRIFT IS GOEDGEKEURD DOOR DE PROMOTOR PROF. IR. J. P. W. HOUTMAN
to my parents to Ineke
THE INVESTIGATIONS DESCRIBED IN THIS THESIS
HAVE BEEN CARRIED O U T IN THE LABORATORIES FOR
INDUSTRIAL DEVELOPMENT OF EUROCHEMIC, MOL,
CONTENTS
TERMINOLOGY
LIST OF SYMBOLS AND ABBREVIATIONS INTRODUCTION
Page
9
11
13
CHAPTER I Processes for final purification and concentration
of plutonium 17
1.1 Introduction 17
1.2 Methods used for final purification and
concen-tration of plutonium 19
1.3 The "TLA-direct precipitation" process 22
Extraction of plutonium and its contaminants by
trilaurylamine 24
Properties of the extractant 24
The nature of extraction by trilaurylamine 26
Extraction coefficients 28
Application of data for the purification of plutonium
from uranium, zirconium + niobium and ruthenium 32
11.5 Chemical damage of the extractant 35
11.6 Conclusions 36
CHAPTER II
ILl
II.2
II.3
II.4
CHAPTER III
III.l
III.2
III.3
CHAPTER IV
IV. 1
IV.2
IV.3
CHAPTER V
V.l
V.2
The purification of plutonium by TLA-SOLVES'
SO (100) in a continuous process
Introduction
Experimental approach
Experimental details
III.4 Discussion of results ...
Decontamination of plutonium from ruthenium
Introduction
Composition and properties of ruthenium
com-plexes in nitrate systems
Experimental work
"Direct precipitation" of tetravalent plutonium
with oxalic acid
Introduction
The solubility of plutonium in the liquid phases
of the system 0.32 M TLA-SVO(100)/H20 —
H N O 3 — H 2 C 2 0 4 / P u ( C 2 0 4 ) 2 - 6 H 2 O
37
37
38
39
43
51
51
51
56
62
62
63
V.3 Kinetics of the "direct precipitation" process . . . . 69 V.4 Sedimentation properties of plutonium ( I V ) oxalate 71
V.5 Experimental work 75 V.6 Discussion of results 81
CHAPTER VI Continuous "direct precipitation" and washing of
plutonium (IV) oxalate 85
VI.1 Introduction 85
VI.2 Description of apparatus 86
V I . 3 Plutonium losses in continuous "direct precipitation" 88 VI.4 Chemical conversion in continuous "direct
precipi-tation" 88 V I . 5 Investigation of process conditions for continuous
"direct precipitation" 89 V I . 6 Experimental results 91
CHAPTER V I I Proposed flow diagram for final purification and
recovery of plutonium on a production scale ... 97
A P P E N D I X I Analytical procedures 101 A P P E N D I X II Chemicals used in the experimental work 104
A P P E N D I X H I Method for calculating individual stage efficiency
factors 106 SUMMARY 107 SAMENVATTING 109 REFERENCES I l l ACKNOWLEDGEMENTS 115
8
T E R M I N O L O G Y
Some of the terminology used in this thesis is listed below. It is based on the terminology of Coleman 50).
Decontamination factor: the concentration ratio of an impurity in a stance, before and after purification of that sub-Stance.
Diluent : liquid, diluting the extracting agent (solvesso-100).
Extract : the (organic) phase containing the solute after extraction.
Extractant : the entire (organic) phase used for extraction.
Extracting agent : the extracting chemical compound (trilauryl-a m i n e ) .
Extraction coefficient : the concentration ratio of a substance in the extract and the raffinate.
Extraction isotherm : equilibrium curve indicating the corresponding concentrations of a component in the extract and the raffinate at constant temperature.
Modifier : an auxiliary substance, usually considered as part of the diluent, added to prevent phase separation of the extract.
Raffinate : the (aqueous) solution separated from the extract after extraction.
Scrubbing : the equilibrating contact of a solution with the extract to remove contaminants.
Separation factor : the ratio of the extraction coefficients of two extractable substances.
Solute : the component to be extracted.
Stripping : the equilibrating contact of a solution with the extract to remove the solute.
L I S T O F S Y M B O L S A N D A B B R E V I A T I O N S
Those symbols and abbreviations which appear in different chapters are listed below:
B total surface of interface between organic a n d aqueous phase (m2).
Ci curies or 3.7 x W^ disintegrations per second. D F w decontamination factor of an impurity M .
E-KT extraction coefficient of a compound M .
F flowrate (m*/sec.)
k the mass transfer coefficient of plutonium across the organic-aqueous interface ( m / s e c ) .
M molarity ( k g m o l e s / m ' ) .
M W d / t thermal megawatt days per ton.
p p m parts per million. r.p.m. revolutions per minute.
S relative sedimentation rate ( m / s e c ) .
Syr separation factor of a component M from plutonium.
S V O (100) Solvesso (100), a mixture of aromates consisting mainly of isomeres of trimethylbenzene (Appendix I I ) . T B P tri-n-butyl phosphate ( G 4 H 9 0 ) 3 P O . T C A tricaprylamine ( C H 2 ) 7 . i i ( C H 3 ) 3 N . T L A trilaurylamine or tridodecylamine ( G i 2 H 2 5 ) 3 N . T O A tri-n-octylamine ( C 8 H i 7 ) N . V volume (m*). P stability constant.
0 the flowrate of a solution ( m ' / s e c ) . [ ] concentration ( k g m o l e s / m » ) . Subscripts
a
L
min.
0susp
t
aqueous phase
liquid
minimum
organic phase
suspension
time
I N T R O D U C T I O N
In December 1955, the Eurochemic company for the chemical reprocessing of irradiated nuclear fuels was founded under the auspices of the European Nuclear Energy Agency and was based on a special convention signed by thirteen European c o u n t r i e s ' ) . T h e reprocessing plant since then built by this company has a multipurpose character enabling a wide variety of nuclear fuels from European reactors to be treated. Although the general flow sheet adopted is of the " P u r e x " type ^) two alternative methods for the final purification of plutonium can be applied with the equipment foreseen*). T h e first is bcised again on the " P u r e x " flow sheet using T B P (tri-n-butyl phosphate) as extracting agent, while the second is based on extraction by T L A (trilaurylamine). T h e latter process is described in this thesis. T h e T B P process has the advantages of being well known and being supported by years of operational experience. T h e T L A process, on the other hand, seems to offer many definite advantages over the T B P process, but still requires operational experience beyond laboratory scale.
T h e work described in this thesis deals with a number of chemical and technological aspects of the development of the T L A process.
T h e chemical flow sheet generally applied for the reprocessing of irradiated fuels can be divided as follows:
1. Dissolution of the fuel.
2. Initial purification of uranium and plutonium from fission products. 3. Separation of plutonium and uranium.
4. Final purification and conditioning of uranium, also called the "uranium tail-end" process.
5. Final purification and conditioning of plutonium, also called the "plutonium tail-end" process.
Figure i. 1 presents the basic Eurochemic flow sheet for the initial purifi-cation of uranium and plutonium from fission products, and the rough separation of these two actinides from each other. This flow sheet has been described by Barendregt^) ^). It shows the paths followed by plutonium and uranium until they lead to the "tail-end" processes.
Before the fuel material is dissolved, the fuel elements are decanned by dissolving the cladding material in a suitable reagent ("head-end" process). T h e dissolving agents used here are sodium hydroxide, sulphuric acid or a m m o n i u m fluoride, depending on the type of cladding material, which m a y be aluminium, magnesium ( m a g n o x ) , stainless steel, zirconium or zircaloy ^ ) .
T h e solution containing the cladding material is stored as medium active waste (j8 + y activity 0 . 1 — 4 Ci/1). T h e fuel material itself, consisting of uranium dioxide or metallic alloys of uranium, is dissolved in 9 M nitric acid. This solution is transferred to the first column of the extraction cycle (Figure i . l ) .
T h e extraction cycle consists basically of five pulse type columns. I n the first two columns, the u r a n i u m and plutonium are both extracted by a mixture of T B P (30 volume % ) and kerosene. In column 1 the greatest p a r t of the fission products is separated from uranium and plutonium. I n the second column additional separation is obtained by scrubbing the T B P phase by an aqueous solution of relatively low nitric acid concentration. T h e aqueous solution leaving the first column is evaporated to reduce the volume of radioactive waste to be stored. I n the third column, the plutonium is stripped from the T B P phase by an aqueous solution containing ferrous sulphamate or u r a n i u m ( I V ) nitrate. These latter agents reduce the plutonium to P u ( I I I ) , which is much less extractable by T B P than P u ( I V ) . As the aqueous solution leaving the third column contains a significant amount of uranium, it is scrubbed by a fresh solution of T B P in column 4. T h e uranium in the organic solution that leaves column 3 is stripped by means of dilute nitric acid in column 5. T h e resulting aqueous solution is transferred to the " u r a n i u m tail-end" cycle (not described here) for final purification and conditioning of uranium. T h e aqueous plutonium solution leaving column 4 also needs further purification and subsequent conditioning to plutonium dioxide. T h e T L A process, subject of this thesis, is part of this "plutonium tail-end" process and involves purification of plutonium by extraction with T L A and precipitation of plutonium as plutonium oxalate from the T L A phase. T h e precipitate is then easily convertible to the oxide by calcination. By precipitating the plutonium directly from the extract a special stripping unit, as used in all existing "tail-end" processes based on solvent extraction, is avoided.
Objectives of the Work
T h e main objectives in the development of the T L A process were: — to achieve sufficient decontamination of plutonium from uranium,
zirconium, niobium and ruthenium,
— to obtain a solid product by precipitating the plutonium from the T L A phase with a suitable precipitating agent. T h e precipitate should be convertible to plutonium dioxide by calcination,
— to develop an apparatus that enables the precipitation process to be carried out continuously.
Arrangement of this thesis
I n C h a p t e r I, the various methods for purification and concentration of plutonium are discussed, as well as the principal characteristics of the T L A process studied. I n C h a p t e r II, some extraction characteristics of plutonium
and uranium in a TLA system are described, while in Chapter III the residts
of continuous liquid-liquid extraction experiments are presented. Chapter IV
is specially devoted to methods for improving the decontamination from
ruthenium, as poor decontamination factors for this element are normally
obtained with extraction processes based on TLA. In Chapter V, a systematic
study of the precipitation of plutonium from the orgzmic extract by means
of oxalic acid is presented. From this study, optimum precipitation conditions,
resulting in minimum losses of plutonium in the mother liquor, are deduced.
The influence of certain process variables on the sedimentation characteristics
of the precipitate, are also discussed in this chapter. The development of an
apparatus for continuous precipitation of plutonium oxalate forms the subject
of Chapter VI. From the data and results of the previous chapters, a final
flow sheet for the purification and concentration of plutonium is proposed in
Chapter VII.
Fuel dissolved in 9M HNO3 Feed adjustment U 1.2 M Pu L O X I Ö ' M HNO3 3 M / * • ƒ activity from fission products 1.1 I 1 0 ' C i / m ' 14 TBP ( 3 0 V o l % l in Kerosene
Aqueous phase stream Organic phase stream F = Flowrate I m ' / s e c ) x 1 0 ^ HNOs 4.0 " 1 u Pu / ' • J ' F = • 0.39 M 1.2 > I Ö ' M activity 15 C i / m ^ 4 6 I HNO3 0.16 M FelNHjSOjli 0.01 M
i
HN03 u Pu /3'~
F 0.5 X 3 M 10" M traces S act 6 . 1 0 -= 2 5 vity C i / m ^ U Pu^ =
i
0 3 7 0.31 X 1 0 ' activity 5 C i / m 4 6 M M HN03 u Pu 2.44 M 0.23 M 0.02 M U HNO, 0.34 M a 0 6 M 5 0 TBP HNO3 (30Vol %) 0.05 M I U HNO3 0.13 M 0.2 M F = 4.1 J T B P ( 3 0 V o l ° o ) Kerosene HNO3 1.0 M Pu"* 6.8 X I Ö ' M U " " 0 5 I I Ö ' M /d -f J' activity ~ 5 0 C i / m ' F = 2 1 HNO3 2 . 4 2 U** 0.27 P u " 6.8x1(5' F ^ 2 1 To 'Plutonium tail-end" To Solvent recovery HNO3 0.01 M U 0.34 M /S * ƒ activity ~ 2 . 5 C i / m ' 5 0 To "Uranium tail-end" 1 Separation of U and Pu from fission products2 Additional separation of U and Pu from fission products 3 Separation of U from Pu
i Additional separation of U from Pu 5 Stripping of U
Figure i.l. Flow diagram of the first extraction cycle at Eurochemic for fuel elements with a burn up of l,0OOMWd/t and a cooling period of 600 days^'')
CHAPTER I
P R O C E S S E S F O R F I N A L P U R I F I C A T I O N A N D C O N C E N T R A T I O N O F P L U T O N I U M
I.l Introduction
In general, the plutonium end-product from a chemical reprocessing plant should be a nuclear grade plutonium dioxide powder, which may either be used directly as a base material for manufacture of nuclear fuels or be converted into plutonium tetrafluoride for the production of plutonium metal.
Therefore, the "plutonium tail-end" cycle generally comprises:
— final purification of the plutonium contained in the aqueous product solution leaving the first cycle of the " P u r e x " process (figure i . l ) ; — concentration of the plutonium by precipitation with oxalic acid
or hydrogen peroxide;
— conversion of the precipitate to plutonium dioxide by a calcination process.
T h e experimental work described in this thesis is limited to the first two steps, i.e. the transformation of the precipitate into the oxide is not included.
I.I.I Type of contaminants
T h e " P u r e x " process has been developed basically to treat irradiated fuel elements containing uranium of low enrichment. In such fuels plutonium will only be present in relatively small amounts. In irradiated natural uranium with a b u m - u p of 1000 M W d / t , the mole ratio U / P u amounts to more than 10».
I n the first cycle of the " P u r e x " process, plutonium is separated from the bulk of the uranium. As figure i.l indicates, the mole ratio U / P u in the plutonium product solution leaving the fourth extraction column still amounts to 0.07. I n order to meet specifications of end-product purity (compare § 1.1.2) a further separation of the plutonium from uranium will therefore be essential.
Radioactive contaminants
T h e main radioactive contaminants produced in irradiated fuel elements are fission products and radionuclides formed from non-fissionable com-ponents (additives, cladding in the fuel material) including their decay products.
I n particular those contaminants which emit g a m m a rays of high energy ( > 0 . 5 Mev) or their precursors should be excluded from the final plutonium product in order to avoid extra problems (shielding) in the safe handling and storage of this final product.
Of these nucUdes '^Zr, *5Nb, »«*Ru and i^'Ru are known to have such high solubility in T B P that they are carried over to column 4 t o such an extent t h a t additional purification of the plutonium solution, leaving this column, is generally required.
T h e radionuclide loeR^ itself does not emit gamma rays (only weak beta rays), b u t its decay product i**Rh (half-life 30 seconds) emits strong g a m m a rays ( o o 0.5 M e v ) . Therefore, lO^Ru must be considered as a serious contaminant. T h e nuclides ^^Zr and '^Nb emit gamma rays of almost equal energy; they are generally considered together (Appendix I ) . Because of practical reasons this thesis is limited to the study of the decontamination of plutonium from ' ^ Z r -t- ^^Nb and from '"^Ru. T h e decontamination d a t a of t h e latter nuclide will of course also be vaUd for i**Ru.
Non-radioactive contaminants
Of the non-fissile material commonly present in the fuel, only zirconium is extracted by T B P to a great extent. Furthermore, the a m o u n t of metallic impurities allowed in the end-product (0.5 weight % ) is relatively high compared with the maximum j>ermissible amount of uranium a n d fission products. T h e elimination of non-radioactive contaminants has therefore not been further considered.
1.1.2 Purity of end-product
Specifications for the final purity of the plutonium end-product are not uniformly accepted, as they depend strongly on the typ» of reaotor in which the plutonium will be used.
Generally, the reprocessing industries, including the Eurochemic plant, try to produce a final plutonium product with the following maximum levels of impurity 21) 22):
—• a total metal impurity (uranium plus any other) of 5000 p p m ; — a specific uranium content of 200 p p m (2x10"* kmoles U / k m o l P u ) ; — a specific g a m m a activity from impurities (y rays > 0.5 Mev) of
8 /iCi per gram plutonium (1.9 C i / k m o l P u ) ;
— a specific beta plus g a m m a activity from impurities of 25 ju.Ci per gram plutonium (6.0 C i / k m o l P u ) .
T h e g a m m a activity specified is of the same order of magnitude as the specific activity of plutonium itself, which varies with the isotopic composition. Therefore, decontamination from g a m m a emitting contaminants below this residual level is generally not justified, as it would not reduce the shielding requirements for further treatments of the end-product.
1.1.3 Decontamination factors
T h e degree of purification is usually expressed by a decontamination
factor D F being the ratio of t h e concentrations of an impurity (per unit weight of recovered product) before and after purification. I n order to estimate the D F required in the final purification process, the chemical composition of the feed solution entering the "plutonium tail-end" process must be known. This composition varies with the original fuel typ)e, the burn-up and with the operating conditions in the previous extraction cycle. Table I.l shows three classes of irradiated uranium fuel from existing reactors *) with different bum-ups.
T A B L E I.l Composition of irradiated fuel elements with different initial 235U content and b u m - u p s after a cooling p>eriod of 150 days * ) .
Initial ^^V content in luel (mol % ) 1.4 0.72 0.72 Burn-up ( M W d / t ) 10,000 1,000 100 Pu content (kg/tonne U ) 4 0.8 0.1
Beta + gamma activity (curies/tonne U )
3 X 10» 4 X 105 3.5 X 10*
In order to meet the final specifications for plutonium stated in § 1.1.2, an overall decontamination factor of m a x i m u m 3 x lO'' (for 10,000 M W d / t ) from beta and g a m m a contaminants has to be reached over the entire recovery process.
T h e decontamination factor required in the "plutonium tail-end" process depends on the decontamination factors obtained in t h e previous main decontamination cycle. I n the first extraction cycle of the Eurochemic process, a decontamination factor for beta and gamma contaminants (DF (p + y)) of about 10* is expected (for fuel with a burn-up of 10,000 M W d / t ) , which requires a fairly high additional decontamination factor (3 x 10») in the "plutonium tail-end" process. For comparable reasons an additional decon-tamination factor of 10* is required for the purification of plutonium from uranium. O n e of the objects of the development of the T L A process described in this thesis was therefore the achievement of high decontamination factors. As will be shown in chapters III and IV, decontamination factors of 10* from fission products and 10* from uranium have been achieved with the T L A process, whereby the final product purity requirements were easily met.
1.2 Methods used for final purification and concentration of plutonium
Very often in fuel reprocessing, purification and concentration are not achieved at the same time in one single operation but are carried out in two or more steps. I t should be noted however that sometimes concentration is obtained during the purification step while also some decontamination is achieved in the concentration step.
*) This includes fuel of experimental power reactors and research reactors in which the life-time of the fuel does not only depend on burn-up.
1.2.1 Purification of plutonium
Methods which have been used so far on a production scale are listed in Table 1.2.
T A B L E 1.2 Production-scale methods for plutonium purification Method Used at
Ion e>:change Hanford'^), Savannah River"). Trombay -•''), Marcoule " ) ,
Solvent extraction Cap de la Hague ^^), Windscale'•'),
Oak Ridge'»), Eurochemic-"), Dounreay'*), Co-precipitation Los Alamos ^•t) (1940—1945).
Of the methods listed in this table the co-precipitation method has only historical value and is not applied anymore. Regarding the other methods, each has its advantages and disadvantages. A prop>er choice depends strongly on the total quantity a n d initial impurity of the plutonium to be purified 2"). In the following description, the aspects of t h e various methods are indicated.
Ion exchange processes
In the U n i t e d S t a t e s ' ' ) , C a n a d a ^ ) ' ' ) and Europe 9) various ion exchange processes have been developed for final purification of plutonium nitrate solutions. Cation exchange processes based on the sorption of trivalent plutonium were among the earliest methods used but were abandoned very soon as the unideal chemical behaviour of the plutonium leads to complicated technical processes i") They have therefore been replaced by processes based on anion exchange, since D u r h a m and Mills >*) demonstrated that the anionic nitrate complexes formed by plutonium in nitric acid solutions are specifically bound by amine-based anion exchange resins such as Dowex-1 and Permutit S K ' ^ ) . W i t h this type of resins, decontamination factors greater than 5x10^ for fission products and greater than 5x10* for uranium and other metallic impurities, have been obtained i^). In many plants they have been replaced by solvent extraction processes * ) .
Solvent extraction
Solvent extraction is mostly applied to nitrate systems, since strongly complexing anions such as sulphate, phosphate, fluoride or oxalate ions tend to keep plutonium in the aqueous phase. Numerous organic solvents have been investigated. Tributyl phosphate (TBP) and tertiary amines have found application on a production scale. Compared with ion exchange methods, solvent extraction processes have been shown to be more suited to continuous *) The principle advantage of ion exchange processes is their highly satisfactory performance in treating more diluted plutonium solutions, such as waste liquids from various chemical processes.
processing. They can be carried out in more compact equipment. Furthermore, solvent extraction requires fewer control devices, which is attractive for construction and operation of leak-tight equipment. T h e method is therefore of particular interest for treating large quantities of plutonium as required in the reprocessing of nuclear fuel.
Purification of plutonium by solvent extraction on a technical scale has been reported by several investigators *3) i6). T B P is the most commonly applied extractant; the plants at Windscale *») and Dounreay ''•*) make use of this solvent.
Amines have recently been proposed as an alternative extracting agent
especially suitable for the plutonium "tail-end" process. High selectivity for tetravalent plutonium when present in nitrate media, resulting in good separation from uranium and fission products, is the main attraction besides chemical and radiolytic stability. A number of processes have been developed so far. They are all based on data originally obtained in the U.S.A. *8), which have been extended considerably by French Workers. A number of varieties have been developed in Italy, Sweden 2') and at Eurochemic^"). T h e most important processes are described hereafter.
T h e Oak Ridge process
This process >") is still in an experimental stage. It is designed to accept a plutonium solution of 0.5—2 g P u / / . T h e plutonium is first extracted by a solution of 0.3M TLAHNO.-j in dimethylbenzene and subsequently scrubbed with a solution of nitric acid. Both low (0.5 M ) and high (5.0 M ) molarity of the nitric acid solution are used. T h e r e has not been sufficient testing with actual plant solutions to establish a preference. T h e T L A phase is subsequently stripped by a mixture of nitric acid and acetic acid. T h e decontamination factors obtained during extraction and stripping is 1 x 10* for uranium and iirconium while the plutonium losses do not exceed 0.2 weight % .
T h e French process
T h e French process also makes use of T L A as extracting agent and is based on the development work carried out by Chesné et al^fi). T h e extractant consists of a mixture of TLAHNO.-j in dodecane and octanol. Extraction is carried out from 2.0 M HNO;^ while the extract is scrubbed vrith an aqueous solution of O . 5 M H N O 3 . T h e plutonium is then stripped from the extract by an aqueous mixture of sulphuric acid (1.5 M ) and nitric acid (0.07 M ) . T h e process is applied in the reprocessing plants at Marcoule *5) and at C a p de la Hague ^s) to complete the decontamination of plutonium which is recovered from irradiated fuel in one or two "Purex" cycles. I n the Mar-coule plant the extractant consist of a mixture of T L A H N O 3 (0.15 M ) , octanol (2-3 vol. % ) and dodecane (§ I I . 1 ) .
In C a p de la Hague the amine content (0.32 M T L A H N O 3 ) is higher, in order to obtain a higher plutonium loading of the extractant.
I n both reprocessing plants an overall decontamination factor of about 1x10* for both u r a n i u m and zirconium, has been achieved (during extraction and stripping), while the plutonium losses did not exceed 0.2 weight % .
1.2.2 Final concentration of plutonium
I n nearly all processes where plutonium is purified by solvent extraction or ion exchange, final concentration of plutonium is accomplished by pre-cipitating it in a solid form that is easily convertible to plutonium dioxide. T h e type of precipitate is selected in accordance with the use intended for the solid compound. Often, a dense nuclear grade sinterable plutonium dioxide powder is required. T w o processes are in current use on a production scale:
— the plutonium p)eroxide process, a n d — the plutonium ( I V ) oxalate process.
I n both processes, plutonium dioxide is obtained by calcining the precipitate at 600°C.—900°G.
Precipitation of plutonium peroxide
Plutonium peroxide (PU2O7) has been satisfactorily precip>itated from feed solutions containing 10—100 g Pu/Z^*). Precipitation is accomplished by addition of H2O2 to t h e plutonium solution, which is between 1 a n d 3 molar in nitric a d d . T h e pwocess is particularly suitable if the precipitate has t o be dissolved subsequently in nitric acid, which readily dissolves plutonium peroxide. T h e process has been applied on a plant scale at Savannah River ^2) and at Los Alamos i^). I n the latter plant it was replaced by the oxalate process after it was found that the peroxide precipitate needed an extra alcohol wash or a slow drying step when the product h a d to be used as base material for hydrofluorination.
Precipitation of plutonium(IV) oxalate
Plutonium has been satisfactorily precipitated with oxalic acid from solutions containing 1—300 g Pu/Z. T h e solubility of plutonium (IV) oxalate is much lower t h a n t h a t of plutonium peroxide in nitric acid solutions. Precipitation at 50—60°C yields products with excellent filtration and sedimentation characteristics. T h e process has been applied in a number of rep>rocessing plants after a purification process which makes use of either ion exchange i ' ) or solvent extraction with either T B P *') or T L A ^^) as extracting agent.
1.3 T h e " T L A — d i r e c t precipitation" process
As this process is extensively discussed in the following chapters, only its basic goals and features are discussed here. T h e development of this process was started in 1958 at Eurochemic to achieve certain advantages i.e.:
— a higher loading capacity of the extractant (approximately 25 g
Pu/Z; (in the French and American processes the maximum loading
capacity varies between 2 and 9 g Pu/Z)
— the elimination of a stripping unit by precipitating plutonium directly
from the extractant.
In principle this is accomplished by:
— using 0.32 M TLA in Solvesso (100) (see appendix III) as extractant.
It will be shown that the use of solvesso (100) as aromatic diluent
results in a considerable increase in loading capacity of the extractant.
— using a combined stripping and precipitation step, caused by a direct
contact between the organic phase and a solution of oxalic acid and
nitric acid. This process step is called "direct precipitation" to
indicate that stripping and precipitation are accomplished
simultane-ously. Such a process needs the development of specific equipment
and the selection of specific process conditions. Part of this thesis
is devoted to this development work.
CHAPTER H
E X T R A C T I O N O F P L U T O N I U M AND I T S C O N T A M I N A N T S BY T R I L A U R Y L A M I N E
TT.l Properties of the extractant
As has been discussed in § 1.2.1, tributylphosphate (TBP) has two major disadvantages as extracting agent for purifying plutonium from uranium and fission products:
— it is not highly selective for plutonium,
— hydrolysis and radiolysis of T B P yields mono- and dibutylphosphates which increase the extraction of fission products, particularly zirconium and ruthenium, resulting in a decrease in the decon-tamination factor.
Considerable effort has therefore been devoted during the last ten years to investigating new extraction systems^*) 2*). Among these systems, those involving tertiary amines show high selectivity for plutonium over a wide range of chemical conditions. Moreover, this selectivity is less vulnerable to chemical and radiation damage of the amines as compared with T B P * " ) . In 1948, Smith and Page»") reported the extraction of mineral acids and negatively charged ion complexes by long-chain tertiary amines. In 1952, Moore »*) used these systems for analytical purposes. This was followed by extensive studies by Coleman »2) »») and Brown»*). T h e tertiary amines containing hydrocarbon groups in the range of 6 to 12 carbon atoms, have been investigated most extensively, as these amine salts show very low solubility in water »5). As already mentioned (chapter I) the use of tertiary amines for the purification of plutonium was started at Oak Ridge in the U.S.A. By comparing different amines trilaurylamine showed excellent pro-porties. In particular these can be described as:
a) its high selectivity in extracting plutonium from nitric acid media over a large range of acidities;
b) its high extracting power at lower acidities as compared with most other amines *^) 36-39);
c) its favourable physical and chemical characteristics such as:
— relatively low density (835 kg/m») providing good separation characteristics from the aqueous phase;
— low solubility in water ( ^1 10 p p m ) resulting in small losses; — relatively high boiling point (224°C at 0.1 m m mercury pressure)
resulting in low losses caused by evaporation;
— presence of hydrocarbon chains of sufficiendy high molecular weight to give good solubility in organic diluents;
— low vulnerability with respect to chemical and radiation damage 35) 38) 4 0 ) .
d) its availability in a pure form and in large quantities.
T h e effects of radiation on the extraction behaviour has been the subject of various investigations 39) 41-43). T L A H N O , ~ SVO(IOO) mixtures had to be exjKJsed to about 2 megarad to observe any effect on solvent properties such as surface tension, retention of uranium and fission products and the tendency of separation of the extractant into two immiscible layers. This dose is far above that encountered in a "plutonium tail-end" process from gamma rays, which is of the order of 10» megarad per cycle*»). T h e influence of a-radiation originating from the plutonium has not received extensive attention in literature. A liquid containing 15 g Pu/Z receives 10"- megarad per hour. As the residence time of such a liquid in t h e extraction equipment is not more than one hour an additional «-radiation dose of 0.01 megarad per extraction cycle can be expected. It may be said however that the damage caused by a-radiation can be considerably higher than the damage inflected by g a m m a radiation. T o make sure that this would not lead to unwanted effects, a Pu-loaded T L A solution (15 g Pu/Z) was stored for some months. No appreciable change in extraction characteristics could be observed. Tertiary amines are generally dissolved in a hydrocarbon diluent to obtain an extractant with suitable physical characteristics like low density, low viscosity and high surface tension at the aqueous interface. Furthermore, the diluent must show good radiation stability and chemical resistance, low vapour pressure at room temp>erature, a reasonably high flash point, low solubility in water and a high ability to dissolve the relevant amine-metal complexes. This last requirement is not easily satisfied, as most amine-diluent mixtures separate into two immiscible layers when loaded with metals of high atomic weight such as plutonium. When for instance plutonium is dissolved in a solution of 0.15 M T L A H N O 3 in dodecane to a concentration of more than 2 grams plutonium per litre, the extract is no longer stable and separates into two layers of different densities **). This so-called "third phase" formation has been a serious drawback to the use of tertiary amines in most extraction systems. Small amounts of a polar solvent (modifier) such as octyl alcohol can be added to aliphatic diluents to reduce the tendency of the organic layer to form a "third phase" 45). However, many modifiers have a negative influence on the extracting power of the amine for plutonium. " T h i r d phase" formation has often been reported, but there is much disagreement in ex-plaining this phenomenon 46-48) Kertes*") recently discussed the information accumulated u p till 1965 and concluded that "third phase" formation is closely related to the tendency of amine salts to associate with each other or with the amine-metal complexes by dipole-dipole interactions. This association results in the formation of aggregates and is more pronounced in aliphatic
hydrocarbons (with generally a relatively low dielectric constant) t h a n in aromatic hydrocarbons. Solvesso (100) (appendix H I ) was chosen as diluent in our studies, as it did not show any phase separation in the system examined, even when the extractant (0.08 — 0.32 M T L A H N O 3 ) was completely saturated with plutonium (see § I I . 3 . 2 ) . For our experimental work a 0.32 M solution of T L A H N O 3 in Solvesso (100) was chosen as extractant so t h a t the results could be compared with those of Coleman ^o), who h a d obtained some d a t a with this extractant at O a k Ridge.
II.2 The nature of extraction by trilaurylamine
Several investigators s*^^) have shown that the extraction of mineral acids and actinides by tertiary amines is based on the formation of neutral com-pounds which are much more soluble in the organic than in the aqueous phase. T h e extraction of mineral acids, and actinides dissolved in these acids, can be described by the following extraction equilibria:
nR3N(o)+mHL(„)^(R3N)n(HL)n,(„)
(R3N)n(HL)m^^)-(-MpLq(,,)^(R3N)nHmMpLm-^q(o) (R3N)nHmMpLm+c|^^,^+rS(^^^(R3N)nHmMpLm+qSr(,
I n these equations R 3 N represents the tertiary amine, H L t h e acid (L = ligand), M the actinide with valence state q, S an associated group of com-pounds present in the organic phase such as the diluent (solvatation) or ooextraoted water or free alkylammonium nitrate molecules. T h e subscripts o and a denote the organic and aqueous phase respectively in which t h e component is dissolved. T h e values for the integers n and p dep>end not only on t h e valencies of the actinide and the ligand but also on t h e typ)e of amine
(length of alkyl group, chain branching) and the typ>e of diluent.
T w o methods are in current use to determine n and p or their ratio: a) T h e method of maximum loading of extractant. This method consists of
determining t h e concentration ratio of actinide and amine in the organic phase when this phase is saturated with the acünide. T h e method has the drawback that it only gives information at saturation and not at lower actinide concentrations.
b) T h e method of applying the mass action law to the extraction equiUbrium
nR3>to)+'"HL^a)+^^pLqCa)=(%N)nHmMpLm^^j I n this case one m a y write:
^(R3N)nHmMpLm-Hq^[(R3N)nHmMpLin_Hq]p ^' (-R3N)" x[R3N]:;x(aHL)'"x[HL]-xaMpL,KMpLci]a I L l
n.2
II.3 II.4 II.526
^^^ [(R.sN)nHmMpLm+q]o
[MpLq]a
is identical to the extraction coefficient E j ^ , equation II.5 can be substituted by:
^^ ^(R3N)nHmMpLm^q ^ ^ M ^^ ^ (^R3N)x[R3N]Sx(^HL)'"x[HL]3xa^pLq
I n these formulas K is the equilibrium constant and a t h e molar activity coefficient of the relevant component. T h e value n could be deduced from the slope of the curve representing the function E ^ == ^[^3^10 at constant acid concentration in the aqueous phase provided the molar activity coefficients are known. Because these coefficients are generally not known, many investigators ^s) 56) have assumed them to be equal to one. As a result some values of n have been published which h a d to be revoked afterwards when it became more and more evident 57-65) t h a t the amine compounds (in particular when dissolved in diluents of low dielectrical constant) have a strong tendency to associate with each other and with other components of the organic phase. T h e degree of association is extensively under investigation at O a k R i d g e " * ) . Here, the molar activity coefficients of the amine compounds in many extraction systems are determined osmotically, wheras others make an a t t e m p t by isopiestic and vapor pressure measure-m e n t s " " ) . Unfortunately, d a t a so far obtained are scarce so t h a t the results found by the two methods cannot be properly evaluated.
A complete study of the extraction mechanism should not only include identification of the components involved, but also investigation of the inter-mediate steps through which the extraction process proceeds. Spectral analyses at intermediate stages are used to investigate whether the extracted compounds are formed in t h e organic or in t h e aqueous phase ^9). I t should be noted however, that the interpretation of the kinetic data in such systems is hampered by the fact that the observed d a t a are also influenced by mass transfer through the organic-aqueous interface (i.e., by type and intensity of mixing).
N o investigator has yet proved a specific extraction mechanism for tertiary amine systems. All mechanisms reported are based on assumptions. V a n Ipenberg ^*) 55) for instance assumes that the amine-metal complexes are formed in the aqueous phase, either in t h e bulk or near to the interface, and are then extracted. His assumption is based on the fact that chemical reactions are generally much faster in the aqueous phase t h a n in the organic phase. Moreover, he believes t h a t the amine which contains a polarisable group is more likely to go into the aqueous phase than the strongly polar metal salts to pass direcdy into the organic phase.
McDowell ajid Coleman 67) recently suggested t h a t the extractable com-pounds are formed at the interface. I t seems indeed very likely t h a t the
hydrocarbon groups of the tertiary amine remain in the organic phase, while the ammonium salt group is attracted by the aqueous side of the organic-aqueous interface. T h e ammonium salt group could then be ionised in the aqueous phase with the result that the — N H + ions combine with the negatively charged ions to form stable neutral amine-metal salts which could then diffuse into the organic phase.
II.3 Extraction coefficients
II.3.1 Literature data Extraction of nitric acid .
Keder, Sheppard and Wilson 56)««) and Baroncelli 53) have investigated the extraction of actinides by tertiary amines from aqueous solutions containing different typ>es of mineral acids. They found that highest extraction was obtained in systems containing nitric acid.
Equation II.7 gives a simplified description of the extracdon equilibrium of nitric acid by T L A .
T L A ^^^ + n H N O , ^^^ % T L A ( H N 0 3 ) n (o) 11-7 T h e mole ratio n can be considerably in excess of one. T h e maximum value
of n depends very much on the type of amine, the diluent involved aiid the acidity in the aqueous phase. Most investigators attribute the fact that n is generally in excess of one to some type of dipole-dipole interaction between the alkylammoniumnitrate and nitric acid**) ®*) *^).
T h e following empirical equation II.8 hcis been established to describe the total amount of nitric acid dissolved in the amine phase as a function of the acidity in the aqueous phase and the amine concentration in the organic phase.
[HNO:,]^^^ = f^'-'N]^^^ -f k[R3N]^^^ X [HN0.3](^^ n . 8 k is a constant and is experimentally determined. I n table I I . 1 the values of k
are shown for some amine systems of slightly different chemical composition. It shows that the k-value obtained for the TLA-SVO(IOO) systems does not differ very much from those obtained with the other systems.
T A B L E I L l k-values for nitric acid in some amine systems Amine phr.-c Nitric acid concentration Value Investigator
in the aqueous phase of k
TLA-SVO(IOO) 0 . 5 — 5 M 0.19 Rolandi et al «s) TCA*-SVO(100) 1 — 6 M 0.20 Baroncelh"») TLA-Toluene 2 — 10 M 0.17 Vaughen and Mason a») * TCA is tri-n-capry/amine
Extraction of actinides
Apart from the limited results obtained by Coleman 50) at a single molarity no quantitative d a t a are available on the extraction of actinides in the system T L A - S V O ( 1 0 0 ) / H N O 3 - H 2 0 . However, some more complete d a t a have been published on somewhat similar extraction systems. Keder and Wilson "") for
instance studied the extraction of plutonium ( I V ) by mixtures of tri-n-octylamine and xylene. They found slopes of the plot log E p y / j v ) versus log amine concentration very close to two at low amine concentrations ( ^ 0.15 M T O A ) and low plutonium concentrations ( ^ lO^M) suggesting the formation of complexes of the formula ( T O A H ) 2 P u ( N 0 3 ) 6 .
Baroncelli et al 58) studied in particular the extraction of tetravalent plutonium by mixtures of tri-n-caprylamine (0.02 — 0.8 M TCA)-xylene and octyl alcohol (0-2 volume % ) . They found slopes varying from 2 to 3 dep)ending on the concentration of octanol in the extractant. Furthermore sp>ectra of t h e extracts showed absorption maxima at 500, 750 and 1050 m/i, indicating that the plutonium in the organic phase is present, at least in part, in the form of the hexanitrato complex. By combining these two data, they concluded that t h e plutonium spjecies in the T C A phase is best described by ( T C A H ) o P u ( N 0 3 ) 6 .
Brothers, H a r t and Mathers^") recently found from spectral studies that tetravalent plutonium forms a series of nitrate complexes in aqueous solutions of different nitric acid concentrations. At an acidity of 4 M H N O 3 the neutral complex P u ( N 0 3 ) 4 predominates. Below 4 M H N O 3 , the positively charged P u ( N 0 3 ) + ions are present in considerable amounts, whereas P u ( N 0 3 ) ^ and Pu(NO,!i)g~" complexes predominate at about 8 M and 11 M H N O 3 respectively. As each complex has its own extraction coefficient, the overall extraction coefficient of plutonium ( I V ) will change with the nitric acid concentration in the system.
With respect to hexavalent uranium, experimental work has been carried out by both Keder et al 56) and by Sato 58). Keder used trioctylamine as extracting agent and found slop>e values (by plotting log E u ( V I ) versus log amine concentration) between 1 and 2.
Sato p>erformed his extraction exp)eriments with trioctylamine and trilauryl-amine and found slop)es of 1.45 and 1.50 resp)ectively. In a later stage however, saturation experiments indicated that a value for n = 2 would be more acceptable indicating that t h e uranium is most likely extracted in the form of the tetranitrato complex ( T L A H ) 2 U 0 2 (NO3 )4. Later this was confirmed by Sato *") by his spectral studies.
T h e extraction d a t a of Keder et al56) 66)^ obtained with 10 volume % T O A in xylene at different acidities are shown in figure I L l and II.2. They show that maximum extraction of tetravalent plutonium occurs from a 4 M H N O 3 solution which according to Brothers et a l ^ " ) , predominandy contains the neutral P u ( N 0 3 ) 4 salt.
Extraction of fission products
As is discussed in § I . l , the principal fission products encountered in a final purification process are zirconium, niobium and ruthenium. As the purification of plutonium is based on selective extraction of plutonium by T L A , all other nuclides should exhibit a low extraction coefficient in this extracdon system. I t has indeed been shown by Coleman »2) t h a t the
extrac-IQ"" 10-1 — I — I 10-1 — r o PLUTONIUM (IV) A URANIUM(IV) [actinide ]_~10 M _ L _ 1 _ 10 10 ~1 1 — I 1 — I — - . o PLUTONIUM (VI) a URANIUM(VI) . -[actinide 1 . - 1 Ö M _ L _ L _L. _L. _ L _ L 0 2 4 6 8 10 12 14
H N O 3 concentration In aqueous phase (M)
Figure H . l . T h e extraction of tetrava-lent actinides from nitric acid solutions by 1 0 ( v o l ) % T O A in xylene 56) 66).
0 2 4 6 8 10 12 14
HNO3 c o n c e n t r a t i o n in aqueous phase (M)
Figure 11.2. The extraction of hexa-valent actinides from nitric acid solutions by 10(vol)% T O A in Xylene 56) 66).
tion coefficient of zirconium (present in tracer amounts) by tertiary amines is below 0.1 in the acidity region 0.5 — 8 M H N O 3 . Vaughen and Mason »«) established extraction isotherms for zirconium and mthenium for the systems 0.32 M T L A — t o l u e n e / H N O s — H 2 O , which are presented in figure I I . 3 . This figure shows t h a t the extraction coefficient of zirconium is indeed very low and increases nearly propordonally with the nitric acid concentration in t h e range from 2 — 8 M H N O 3 . From the experience obtained with T B P , one is inclined to believe that niobium follows
ziconium in its extraction behaviour 55). In general it m a y be said that extraction of Z r (-I- N b ) b u t especially of R u by T L A shows poor reproducibility. For R u this is caused by its sp>ecific chemical behaviour in « nitrate systems, i.e. by the low formation rates | of the various nitrato complexes involved. <=-T h e effects have caused many investigators ° to come to conclusions which app>eared * 10^ later somewhat speculative a n d incomplete. ^ Ruthenium is considered today to be a really " troublesome fission product in most purifi- .S cation processes based on solvent extraction. 2 10 Special attention is devoted t o this subject m in chapters I I I and I V .
Figure U.S. Extraction coefficients obtained by Vaughen and Mason 56) 66) for zirconium and ruthenium, in the system TLA(0.32 M ) — t o l u e n e / H2O—HNO3.
30
b 1 ^ " ^-1 -1 -1 -1 -1 0 RUTHENIUM (III) A ZIRCONIUM (IV)
\
>^^^_^
/
/
/
V 1 1 1 1 1 q•J
\
!
• :'
0 2 4 6 8 10 12 14II.3.2 Experimental data *
T h e scarcely reported d a t a i * ) » ^ ) on the extraction of plutonium and uranium in the system chosen for our process, m a d e it necessary to carry out some additional extraction exp)eriments. This work comprised measurement of the distribution of plutonium ( I V ) and u r a n i u m ( V I ) in the systems T L A -SVO(IOO) / H N O 3 — H 2 O at room temperature and at various nitric acid and T L A concentrations. Determination of the extraction coefficients of the fission products zirconium, niobium and ruthenium was not considered im-portant as the extraction behaviour of these nuclides in a continuous and multistage extraction system can not be predicted from such data with sufficient reliability (see chapters I I I and I V ) .
T h e plutonium and uranium extractions were carried out in a single stage extractor. T h e actinide nitrate was dissolved in aqueous nitric acid solutions of various concentrations. Exjual volumes (15 ml) of the organic and aqueous phase were mixed for 5 minutes at 23°C with a stirrer at 1000 r.p.m. Stirring for more than 5 minutes was accepted to be sufficient, as Rolandi ®^) had shown that under nearly identical exp)erimental conditions, extraction equili-brium is reached within 1 minute. I n order to prevent a change in the acidity of the aqueous phase by contacting this phase with the extractant, t h e organic phase was equilibrated prior to extraction with a pure nitric acid solution of the same acidity as the actinide solution. After extraction, the two liquid phases were separated by centrifuging and suitable aliquots were then taken for ancdysis. Plutonium was obtained from a stock solution of plutonium ( I V ) nitrate and was kept in the tetravalent state by making the solution 0.05 M in sodium nitrite. T h e uranium solution was prepared by dissolving hexahydrated u r a n i u m ( V I ) nitrate in a nitric acid solution of desired molarity.
T h e extraction coefficients for tetravalent plutonium measured as a function of nitric acid concentration in the system T L A ( 0 . 3 2 M ) — S V O ( I O O ) / H N O 3 — H 2 O are presented in figure II.4. This figure shows curves for relatively low and high plutonium loadings. In figure II.5, the influence of the plutonium concentration in the organic phase on the extraction coeffi-cient is shown. It clearly indicates that saturation of the organic phase is reached at plutonium concentrations in the aqueous phase of about 0.2 M. Saturation values of the plutonium loading in the extractant, have been determined for different amine concentrations and different concentrations of nitric acid. These d a t a are presented in Table II.2.
T h e extraction isotherms of uranium ( V I ) at moderate uranium concen-tration (0.5 —• 2.5xlO'^M) and at different acidities are presented in figure II.6. Here no saturation effects were observed due to the relatively low u r a n i u m concentrations in the aqueous phase. T h e extraction coefficients, derived from the slopes of the curves in figure II.6, are plotted against the nitric acid concentration in figure II.7.
T A B L E II.2. Saturation values for plutonium concentration in the organic phase at different acidities and amine concentrations in the system T L A — S V O ( 1 0 0 ) / P u * + — H N O 3 — H 2 O at 25 °C. TLA ( M ) 0.32 M 0.16 M 0.08 M
0.5 M HNOo in aqueous phase
Plutonium
con-centration at Mole ratio saturation of the T L A / P u
organic phase
0.105 M 3.0 0.050 M 3.2 0.026 M 3.1
2.0 M HNO., in aqueous phase
Plutonium
con-centration at Mole ratio saturation of the T L A / P u organic phase
0.105 M 3.0
Figure II.4 shows that in our system plutonium ( I V ) is maximum extract-able from solutions of approximately 4 M nitric acid. This corresponds to the results obtained by K e d e r et al56) 57) ^vith T O A . Figure II.4 also shows that the limited data of Brown and Coleman for T L A (0.32 M ) ^ S V O ( I O O ) at corresponding plutonium concentration fit our curves very well.
Table II.2 shows that the observed stoechiometric mole ratio of T L A and plutonium, when the extractant is saturated with plutonium, varies between 3.0 and 3.2 for the different amine concentrations investigated. This finding indicates t h a t at maximum plutonium loading, each plutonium ion is associated with three rather than two T L A molecules of the extract. This seems not to be consistent with the conclusions of other investigators as discussed in § II.3.1. It may indicate that association compounds between the extracted plutonium complex and free amine are formed in t h e organic phase. T h e degree of formation of such association complexes will of course depend on the type of amine and diluent used.
Neither the identity of the associated compounds nor their structural configuration are examined any further. O u r extraction curves for hexavalent uranium presented in the figures II.6 and II.7 indicate t h a t the extraction coefficient for uranium is independent of the uranium concentration, at least within the concentration region studied (1x10'» to 2x10"^ M u r a n i u m ) . T h e extraction of uranium is found to increase with increasing nitric acid con-centration in the acidity region between 0.5 and 5.0 M H N O 3 . This is in agreement with the results given in literature for T O A in Xylene (compare figure II.2).
II.4 Application of data for the purification of plutonium from uranium, zirconium + niobium and ruthenium
T h e extraction curves of p l u t o n i u m ( I V ) (figure II.4) and u r a n i u m ( V I )
•1 • /
- /
• 1 — 1 1 1 1 1 : / ^ [Pu]o = 4.1Ö^M [Pu]o»63.1Ö^M-^
'
-
-
. :
O Our data i Data of Brown (16.39) -1 i ..-J -1 L ^ 0 1 2 3 4 5 6 =HNO3 concentration in aqueous phase (M) ^ u
Figure II.4. The extraction of tetra- 0 valent plutonium in the system T L A E °°
(0.32 M ) — S V O ( 1 0 0 ) / H 2 O — H N O 3 at | 25°C. 3 - I T I • 1 I 1 1 I 1 ' / ^ [ H N O a ] ^ - O . S M
/
/ ...
/
[
/
0.05-aoi 1 1 1 1 1 1 1 1 •
^^''
^ °
O ' ' ' ' ^ [ H N 0 3 ] , . 2 . 0 M . / / [ H N 0 3 ] . - 0 . 5 M 1 1 1 1 1 1 1 1 0 0005 1 25°C 1^
-_
1
1 o.c 1 1-~
1 1 0 0.1 02 Plutonium concentration in aqueous phase (IVI)Figure H.S. Extraction isotherms of tetravalent plu-tonium in the system TLA(0.32 M ) ^ S V O ( 1 0 0 ) / H 2 O - H N O 3 .
0 0.005 0.01 0.015 Uranium concentration in aqueous phase (M)
Figure n . 6 . Extraction isotherms of hexavalent uranium in the system T L A
(0.32 M ) — S V O ( 1 0 0 ) / H 2 O — H N O 3 at 25°C.
0 1 2 3 4 5 6
HNO3 concentration in aqueous phase (M)
Figure H.?. The extraction of hexava-lent uranium in the system TLA(0.32 M) — S V O ( 1 0 0 ) / H 2 O — H N O 3 (data dedu-ced from figure I I . 6 ) .
(figure II.7) are combined in figure II.8. O n the ordinate is plotted the separation factor, which is defined by S^p^^.^^^^^^ = ^ — , and on die
abscissa the nitric acid concentration in t h e aqueous phase. For the calculation of t h e separation factor it is assumed here that mutual influence on the extractability of uranium and plutonium is absent. As the extraction coeffi-cient for plutonium depends on the plutonium concentration, two curves for different plutonium loadings are presented in figure I I . 8 . These curves show that the highest separation factor is attained at low acid concentration.
Decontamination of plutonium from zirconium is also achieved at low acidity as indicated by figure I I . 3 . T h o u g h decontamination from ruthenium is highest at relatively high acidity t h e purification at low acidity seems still to be sufficiently high.
T h e limited separation factors observed make a multistage extraction process unavoidable for achieving the high decontamination factors required for uranium ( o o 1x10*) and for fission products (c^:> 3x10^) as discussed in § 1.1.3. Such a process could comprise first extraction of plutonium by t h e T L A phase, followed by scrubbing t h e extract for removal of coextracted impurities; a technique which is followed in most existing T L A processes (chapter I ) .
T h e optimum acidity for the extraction section, obviously depends on the relative amounts of contaminants present in the plutonium feed solution. If the plutonium has to be purified predominantly from zirconium, a lower acidity should be used than if it has to be decontaminated mainly from ruthenium. I n the latter case, the acidity should be as high as possible. However, acidities higher than 5 M should be avoided as chemical damage of the diluent, causing changes in the extraction behaviour, would otherwise become significant (see § I I . 5 ) . I n accordance vrith most existing T L A processes an acidity of 2 M H N O 3 was selected in the extraction section of the process described in this thesis. For the scrub section similar arguments are expected to be valid. Indeed, it was found from t h e results of some preliminary studies that scrubbing the extract with nitric acid solutions is generally very useful for removing coextracted uranium from the amine phase. Contradictory to expectations, this was found not to be the case for the fission products zirconium and ruthenium. T h e removal of zirconium a n d ruthenium from the T L A phase by scrubbing with nitric acid was found to be practically negligible, indicating a m u c h higher retention of these nuclides in the extractant t h a n should be predicted from the extraction d a t a of Vaughen and Mason ^*) ®^). T h e apparent anomalous extraction behaviour of these fission products has already been mentioned in § II.3.1 and wul further be treated in chapters I I I and I V . I n order to achieve maximvim removal of u r a n i u m from the T L A phase, the acidity in the scrub section was chosen to be 0.5 M H N O 3 which corresponds to the lowest acidity permissible in
aqueous solutions containing tetravalent plutonium. At lower acidity, poly-merisation reactions of plutonium are highly p r o b a b l e ' ' ) ''^).
UJ UJ
*b 10' r [Pu]„ = 4x1Ó^M,
[Pu]^= 63»l5^IM.
0 1 2 3 4 5 6
HNO3 concentration in aqueous phase (M)
Figure n . 8 . Separation factors of plu-tonium from uranium in the system T L A (0.32 M ) — S V O ( 1 0 0 ) / H 2 O — H N O 3 at 25°C. (lxIO-3 M < [U]o < 2x10-2) II.5 Chemical damage of the extractant
I n the course of our experiments, it appeared that nitric acid coidd cause chemical damage to the extractant. Aqueous solutions above 7 M H N O 3 changed the TLA-Solvesso(lOO) phase from almost colourless to deep brown. I n order to determine which of the two components undergoes chemical damage, t h e pure components were contacted with nitric acid solutions of different acidities. T h e intensity of the colouring is presented in Table II.3.
T A B L E II.3 Colouring effect of nitric acid on the extractants components (contact time: two days at room t e m p e r a t u r e ) .
H N O 3 concen-tration in aqueous phase (M) 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 TLA pure no colouring SVO(IOO) pure light yellovk' Ï )
»
»
„
J» colouring starts brown colour 0.32 M TLA in SVO (100) light yellovyr»
»)
»
j j JÏ colouring starts broviJn colourThis table shows t h a t the observed colouring results from the aromatic diluent Solvesso (100) rather than from T L A . O u r opinion t h a t this effect has been caused by chemical damage of the aromatic compound has recently
been endorsed by Baroncelli and Grossi, who studied the chemical damage of Solvesso (100) by nitric acid at elevated temperatures. T h e y found t h a t the methylethylbenzenes, the diethylbenzenes and more generally t h e C10H14 isomers present in Solvesso (100) react most easily with nitric acid. T h e y further found that the products formed between 1,2,4-trimethylbenzene and nitric acid exhibit high retention of zirconium upwn stripping with 1 M sodium carbonate. I t seems therefore useful that some future work should be devoted to studying the chemical attack of nitric acid on TLA-SVO(IOO) mixtures, particularly at t h e extraction process conditions, and its effect on the extractability of actinides and fission products.
11.6 Conclusions
T h e experimental data on the extraction behaviour of plutonium and uranium, combined with the reported data on the extractability of zirconium and ruthenium by tertiary amines indicate that purification of plutonium from uranium, zirconium and ruthenium should be possible in a multistage extraction process, using as extractant a mixture of 0.32 M T L A H N O 3 in Solvesso (100). T h e optimum chemical conditions in such a process will depend on the relative concentrations of the contaminants.
CHAPTER III
T H E P U R I F I C A T I O N O F P L U T O N I U M BY TLA-SOLVESSO(IOO) I N A C O N T I N U O U S P R O C E S S
I I I . l Introduction
In chapter I I a number of extraction data have been given for plutonium and its main contaminants present in feed solutions for purification processes based on extraction with T L A . O n e would expect such data to be sufficient for the procedure of designing a continuous process. However, it was found t h a t the anomalous properties of zirconium and ruthenium mentioned in § II.3.1 interfere with such a procedure as these nuclides are more strongly retained in the organic phase than expected from the equilibrium data.
According to Connick and McVey'"*) and Ahrland ^s) the anomahes observed are caused by the fact that zirconium ( + niobium) and ruthenium form various complexes in nitrate solutions. For the case of ruthenium the presence of a series of nitrosyl complexes have been demonstrated. They have been shown to exhibit different solubilities in the extractant **). For zirconium a n d niobium, S i d a i n * ) postulates that the anomalous extraction behaviour is caused by interaction with colloidal material present in fuel solutions. This colloidal material is assumed to form very stable combinations with Z r 4- N b which probably are present in the form of nitrate complexes. These combinations may have high solubility in organic solvents. Sidall does not specify, however, the nature of the colloids that might be involved. O t h e r investigators '3) 77) 78) state that zirconium and ruthenium form stable species with certain impurities present in the extractant and with degradation products formed from the extractant by chemical or radiation damage. Baroncelli and Grossi ^*) recently reported that hydroxamic acids formed by radiation-induced nitration of the diluent in T L A H N O 3 - S V O ( 1 0 0 ) systems are mainly resfx>nsible for the retention of zirconium in the organic phase.
All these explanations involve the presence of zirconium and ruthenium in different chemical forms of which each could display a different extraction coefficient. T h u s one could expect that each species is extracted in a different way in the course of a continuous process leading to a continuous change in the overall extraction coefficient. In addition, the efficiency of a continuous process will depend on the rates of the chemical conversions and physical transport processes occurring during extraction. It has been observed t h a t most of the ruthenium complexes in T B P and T L A nitrate systems display an extremely low rate of conversion to the solvent extractable form^*). T h e kinetics of the physical transport processes can be influenced by proper choice of process equipment.
T h e phenomena observed so far are not sufficient to treat all these details on a quantitative basis. Therefore it was necessary to make use of an experi-mental evaluation of process variables. T h e research which will be described in this chapter includes the study of:
— t h e decontamination of plutonium from uranium, zirconium, niobium, and ruthenium for different feed rates, and
— limitations in use (recycling) of the extractant as a result of radiation or chemical damage.
This investigation was carried out with the use of mixer-settler equipment.
III.2 Experimental approach
T h e process scheme used for most of the multistage extraction experiments is presented in figure I I I . l . For the plutonium-feed solution, a composition was selected in accordance with plant practice. As the plutonium in this solution was present in the trivalent state, it had first to be oxidised to the tetravalent state ( p l u t o n i u m ( I V ) is most extractable by T L A H N O 3 ) by addition of sodium nitrite'^8). T h e plutonium was then extracted in counter-current contact with 0.32 M T L A H N O 3 — S V O ( I O O ) . T h e acidities selected for extraction and scrubbing were based on the data presented in C h a p t e r I I . T h e 2 M nitric acid was chosen for extraction, as good separation of plutonium from uranium and also from zirconium, niobium and ruthenium could be expected (figure II.3 and II.8) at this acidity. T h e extraction coefficient of plutonium ( I V ) being 300 at 2 M H N O 3 (figure II.4) implies that, u n d e r the extraction conditions selected, two theoretical stages would be sufficient to bring the plutonium losses in the raffinate to a value below 0 . 0 1 % . As will be discussed later these 2 theoretical stages have to be replaced by 8 practical stages as the stage efficiency in this part of the mixer-settler bank is rather low.
For the scrub section 0.5 M H N O 3 was chosen with the object to achieve optimum re-extraction of uranium and zirconium from the plutonium-loaded organic phase (compare figures I I . 3 and I I . 8 ) . Although figure II.7 shows that the extraction coefficient of hexavalent uranium decreases with decreasing acidity an acidity lower than 0.5 M H N O 3 was not used, as plutonium has the tendency t o hydrolyse and jx>lymerise in media of lower acidity'^*). I n order to obtain an acidity of 2 M H N O 3 in the extraction section, the feed solution had to be adjusted to 3.5 M H N O 3 . I n one run the flow rate of the plutonium-feed solution was increased to 300 m l / h r . I n this case the feed solution was adjusted to 2.5 M H N O 3 .
Figure H I . 2 shows the McCabe-Thiele diagram for uranium distribution over several stages under the process conditions described in figure I I I . l . T h e figure shows that 6 theoretical stages are required in the scrub section to attain a plutonium product containing less than 200 ppm uranium, which corresponds to 0.4 x 10"* M uranium in the extract solution. This is in accordance with the requirements for the final U-content in the purified
