Immobilization of actinide wastes in ceramics for long-term disposal

204 Pages·2006·9.7 MB·English

Immobilization of actinide wastes in ceramics
for long-term disposal
Von der Fakult(cid:4)at fu(cid:4)r Georessourcen und Materialtechnik der
Rheinisch-Westf(cid:4)alischen Technischen Hochschule Aachen
zur Erlangung des akademischen Grades eines
Doktors der Naturwissenschaften
genehmigte Dissertation
vorgelegt von der Kernphysikerin MSc
Khin Soe Soe
aus Yangon , Myanmar
Berichter: Univ.- Prof. Dr. rer. nat. Reinhard Odoj
Univ.- Prof. Dr. rer. nat. Reinhard Conradt
Tag der mu(cid:4)ndlichen Pru(cid:4)fung: 09. M(cid:4)arz 2006
(Diese Dissertation ist auf den Internetseiten der Hochschulbibliothek online verfu(cid:4)gbar)
Contents
Abstract 5
Abstract 6
1 Introduction 7
1.1 Role of nuclear energy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
1.2 Current management of nuclear spent fuel cycles . . . . . . . . . . . . . . 7
1.2.1 Actinides, (cid:28)ssion products and radiotoxicity . . . . . . . . . . . . . 7
1.2.2 Partitioning-transmutation for actinides . . . . . . . . . . . . . . . 8
1.3 Disposal or transmutation . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
1.3.1 Transmutation stragety . . . . . . . . . . . . . . . . . . . . . . . . . 12
1.3.2 Transmutation technology . . . . . . . . . . . . . . . . . . . . . . . 13
1.3.3 Conditioning for deep geological disposal . . . . . . . . . . . . . . . 18
1.3.4 Stability of the matrix . . . . . . . . . . . . . . . . . . . . . . . . . 19
1.4 Ceramics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
1.4.1 Properties of ceramics . . . . . . . . . . . . . . . . . . . . . . . . . 22
1.4.2 Irradiation damages on ceramics . . . . . . . . . . . . . . . . . . . . 27
1.4.3 Di(cid:27)erent ceramic compounds for waste immobilization . . . . . . . 31
1.5 Zirconium oxides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
1.5.1 Fully stabilized zirconia (or) cubic zirconia . . . . . . . . . . . . . . 36
1.5.2 Phase stability of cubic zirconia . . . . . . . . . . . . . . . . . . . . 36
1.5.3 Simulations elements for incorporation in cubic zirconia . . . . . . . 39
1.6 Properties of zirconia ceramics . . . . . . . . . . . . . . . . . . . . . . . . . 39
1.6.1 Chemical durability of zirconolite and titanate ceramics . . . . . . . 39
1.6.2 Chemical durability of zirconium bearing waste forms . . . . . . . . 40
2 General preparation and characterization methods 45
2.1 Fabrication methods for the ceramic powders . . . . . . . . . . . . . . . . . 45
2.1.1 Sol-gel method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46
2.1.2 Solid state methods . . . . . . . . . . . . . . . . . . . . . . . . . . . 46
2.1.3 Glycerol method . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46
2.1.4 Co-precipitation method . . . . . . . . . . . . . . . . . . . . . . . . 46
2.2 Calcinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47
2.3 Compaction of the nanoceramic powder . . . . . . . . . . . . . . . . . . . . 47
2.4 Sintering of nanoceramic solid materials . . . . . . . . . . . . . . . . . . . 47
2.4.1 Sintering processes . . . . . . . . . . . . . . . . . . . . . . . . . . . 48
2.5 Methodologies for the analysis of powder . . . . . . . . . . . . . . . . . . . 49
XRD
2.5.1 X-ray powder di(cid:27)raction method ( ) . . . . . . . . . . . . . . . 49
BET
2.5.2 Brunauer-Emmett-Teller ( ) method . . . . . . . . . . . . . . . 50
2.5.3 Di(cid:27)erential thermal analysis . . . . . . . . . . . . . . . . . . . . . . 51
2.6 Characterizations of nanocrystalline ceramic solid materials . . . . . . . . . 51
1
2.6.1 Density measurement . . . . . . . . . . . . . . . . . . . . . . . . . . 51
2.6.2 Microstructure in nanocrystalline materials . . . . . . . . . . . . . . 51
2.7 Di(cid:27)erent leaching methods for conditioned radioactive waste . . . . . . . . 52
2.7.1 Basic worldwide leaching methods . . . . . . . . . . . . . . . . . . . 53
2.7.2 Experimental set-up of static and dynamic conditions . . . . . . . . 53
3 Experimental procedures 55
3.1 Materials and apparatus . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55
3.1.1 Applied materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55
3.1.2 Apparatus facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . 56
3.2 The powder for the ceramics . . . . . . . . . . . . . . . . . . . . . . . . . . 57
3.2.1 Synthesis of matrix and ceramic powders . . . . . . . . . . . . . . . 57
3.2.2 Drying . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58
3.2.3 Calcination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59
3.2.4 Grinding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59
3.3 Characterization of the synthesized powders . . . . . . . . . . . . . . . . . 59
3.3.1 The morphology of powder . . . . . . . . . . . . . . . . . . . . . . . 59
3.3.2 Thermal behavior of ceramic powder . . . . . . . . . . . . . . . . . 59
3.3.3 The speci(cid:28)c surface area of the powder . . . . . . . . . . . . . . . . 59
XRD
3.3.4 Crystallographic study of ceramic powders using . . . . . . . 60
X
3.3.5 Structural formation using synchrotron -ray . . . . . . . . . . . . 61
3.3.6 Compaction of ceramic powders . . . . . . . . . . . . . . . . . . . . 61
3.3.7 Sintering of the pellets . . . . . . . . . . . . . . . . . . . . . . . . . 62
3.3.8 Density measurements . . . . . . . . . . . . . . . . . . . . . . . . . 63
3.4 Characterization of synthesized pellets before leaching . . . . . . . . . . . . 63
3.4.1 Polishing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63
3.4.2 Washing and thermal etching . . . . . . . . . . . . . . . . . . . . . 64
SEM/EDX
3.4.3 Microstructural assessments using . . . . . . . . . . . 64
XRD
3.4.4 Characterization of pellets before leaching experiment by . . 64
3.5 Mechanical properties of fabricated ceramics . . . . . . . . . . . . . . . . . 65
3.6 Leaching procedures for ceramics . . . . . . . . . . . . . . . . . . . . . . . 65
3.6.1 Selected methods for leaching experiments . . . . . . . . . . . . . . 65
3.6.2 Removing the sample . . . . . . . . . . . . . . . . . . . . . . . . . . 66
3.6.3 Evaluation of the leaching rate for ceramics containing ceria . . . . 66
3.7 Analysis of the leached sample after the chemical durability test . . . . . . 67
4 Experimental results for yttria-stabilized zirconia ceramics with ceria 69
4.1 Powder after drying . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69
4.2 Morphology of the yttria-stabilized zirconia matrix powder . . . . . . . . . 69
(Zr Y )O 110oC
4.2.1 Morphology of the 0.84 0.16 x powder dried at . . . . . . 69
(Zr Y )O 600oC
4.2.2 Morphology of the 0.84 0.16 x matrix powder calcined at 69
1,100oC
4.2.3 Morphology of the matrix powder after calcination at . . . 70
TG/DSC
4.3 Characterization of ceramic powder containing ceria using . . . 71
(Zr,Y,Ce )O XRD
4.4 Characterization of a x ceramic powder using . . . . . . 72
4.5 Investigations of the properties of ceramics containing ceria before leaching 75
2
4.5.1 Compressibility of the matrix with pressure at di(cid:27)erent calcination
temperatures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75
4.5.2 Compressibility of the matrix with di(cid:27)erent ceria contents . . . . . 76
4.5.3 Microhardness investigation . . . . . . . . . . . . . . . . . . . . . . 78
4.5.4 Structure formation and phase stability of ceria ceramic pellets be-
XRD
fore leaching experiment using . . . . . . . . . . . . . . . . . 79
(Zr,Y,Ce )O
4.5.5 Microstructural assessment of the a x pellets . . . . . . 80
4.6 Investigation of the chemical properties for the ceramics containing ceria . 82
(Zr Y )O
4.6.1 Leaching behavior of 1−x x 2−0.5x matrix pellets under static
conditions in ultra pure water . . . . . . . . . . . . . . . . . . . . . 84
(Zr,Y,Ce)O
4.6.2 Leaching behavior of x ceramic pellets with 5 wt% ceria 85
(Zr,Y,Ce)O
4.6.3 Leaching behavior of x ceramic pellets with 10 wt%
ceria under static condition . . . . . . . . . . . . . . . . . . . . . . 88
(Zr,Y,Ce)O
4.6.4 Leaching behaviour of x ceramic pellets with 10 wt%
110oC
ceria made from powder dried at . . . . . . . . . . . . . . . . 88
(Zr,Y,Ce)O
4.6.5 Leaching behaviour of x ceramic pellets with 10 wt %
600oC
ceria made from powder calcined at . . . . . . . . . . . . . . 91
(Zr,Y,Ce)O
4.6.6 Calcinationtemperaturee(cid:27)ectintheleachingbehaviourof x
ceramics with 10 wt % ceria . . . . . . . . . . . . . . . . . . . . . . 95
(Zr,Y,Ce)O
4.6.7 Thee(cid:27)ectofdensityinleachingbehaviourof x ceramics
with 10 wt% ceria . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97
(Zr,Y,Ce)O
4.6.8 Thee(cid:27)ectoffabricationprocessontheleachingbehaviorof x
ceramic pellets with 25 wt% ceria prepared from powder calcined
600oC
at . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 98
4.6.9 The e(cid:27)ect of ceria content on the leaching behaviour of the pellets
under static conditions . . . . . . . . . . . . . . . . . . . . . . . . . 99
(Zr,Y,Ce )O
4.6.10 Thee(cid:27)ectofceriacontentontheleachingbehaviourof a x
pellets under dynamic test-conditions . . . . . . . . . . . . . . . . . 103
(Zr,Y,Ce)O
4.6.11 Leaching results from x ceramic pellets with 20 wt %
ceria in dynamic leaching test . . . . . . . . . . . . . . . . . . . . . 105
4.7 Characterization of leached samples after the leaching experiment . . . . . 107
(Zr,Y,Ce)O
4.7.1 Microstructure changes of x pellets containing 10 wt
600oC
% ceria prepared from powder calcined at after the leaching
experiment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107
(Zr,Y,Ce)O
4.7.2 Phase stability of x pellets with 10 wt % ceria made
600oC
from powder calcined at after the leaching experiment . . . . 108
(Zr,Y,Ce)O
4.7.3 Phase stability of x ceramic pellets with 10 wt % ceria
110oC
made from powder dried at after leaching . . . . . . . . . . . 110
4.8 Discussion on ceria ceramics . . . . . . . . . . . . . . . . . . . . . . . . . . 119
4.8.1 Physical properties . . . . . . . . . . . . . . . . . . . . . . . . . . . 119
4.8.2 Chemical properties . . . . . . . . . . . . . . . . . . . . . . . . . . . 121
4.9 Conclusions on ceria ceramics . . . . . . . . . . . . . . . . . . . . . . . . . 125
5 ExperimentalResultsforyttria-stabilizedzirconiamatrixwithneodymia131
5.1 Synthesis of neodymia ceramic powder . . . . . . . . . . . . . . . . . . . . 131
5.1.1 Drying and calcination . . . . . . . . . . . . . . . . . . . . . . . . . 131
3
5.1.2 Grinding and the chemical e(cid:27)ect during grinding . . . . . . . . . . 131
5.2 Characterization of the synthesized powder . . . . . . . . . . . . . . . . . . 132
5.2.1 Single phase powders . . . . . . . . . . . . . . . . . . . . . . . . . . 132
5.2.2 Double phase powders . . . . . . . . . . . . . . . . . . . . . . . . . 135
BET
5.2.3 Characterization of powder by . . . . . . . . . . . . . . . . . 139
(Zr,Y,Nd )O XRD
5.2.4 Characterization of powders a x by . . . . . . . . 141
5.3 Properties of ceramics containing neodymia . . . . . . . . . . . . . . . . . 147
5.3.1 Green density and compressibility of ceramics containing neodymia 147
5.3.2 Sinterability of yttria-stabilized zirconia matrix with neodymia . . . 148
5.3.3 Sintered density and the initial properties of ceramic powder . . . . 151
5.4 Characterization of synthesized pellets before leaching . . . . . . . . . . . . 155
5.4.1 Preparation steps before the leaching experiment . . . . . . . . . . 155
(Zr,Y,Nd)O
5.4.2 Microstructure assessment of x pellets . . . . . . . . . 155
5.4.3 Characterization of crystallinity . . . . . . . . . . . . . . . . . . . . 159
5.5 Mechanical properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 161
5.6 Chemical properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 163
5.6.1 Leaching of neodymia pellets in di(cid:27)erent leaching media . . . . . . 164
5.6.2 Leaching in ultra-pure water . . . . . . . . . . . . . . . . . . . . . . 164
5.6.3 Leaching in Mount-Terri-water . . . . . . . . . . . . . . . . . . . . . 168
5.6.4 Leaching in granitic water . . . . . . . . . . . . . . . . . . . . . . . 172
5.7 Conclusions for yttria-stabilized zirconia with di(cid:27)erent neodymia . . . . . . 178
5.8 Perspectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 184
Acknowledgements 185
References 187
Curriculum Vitae 201
4
Abstract
Glas kann mit Zeit entglasen. Der Einbau von Aktiniden fu(cid:4)hrt aufgrund von deren
MeV 106
Alphazerfall im -Bereich im Laufe von mehreren Jahren Endlagerung zu einer
β
zus(cid:4)atzlichen Ausdehnung und Umwandlung von Atommu(cid:4)ll-Glaskokillen. Au erdem ist
die maximale m(cid:4)ogliche Menge der Aktiniden, die in einem Glas enthalten sein kann,
ungef(cid:4)ahr 5%. Das Ziel dieser Dissertation ist die Entwicklung einer Keramik fu(cid:4)r die
Immobilisierung von radioaktiven Abf(cid:4)allen.
(ZrY)O
Yttrium stabilisierte Zirkoniumkeramiken x sind ein vielversprechendes Mate-
rial als Matrix sowohl fu(cid:4)r nuklearen Brennsto(cid:27) als auch fu(cid:4)r radioaktiven Abfall. Grund
hierfu(cid:4)r ist ihre hohe Aufnahmef(cid:4)ahigkeit von Aktiniden, ihre hohe chemische Stabilit(cid:4)at
β
sowie ihre au ergew(cid:4)ohnlich hohe Strahlungsstabilit(cid:4)at u(cid:4)ber eine betr(cid:4)achtliche Temper-
aturspanne. Kubische Materialien expandieren isotrop unter unterschiedlichen Belastun-
gen wenn sie durch Strahlung gesch(cid:4)adigt werden. Dieses fu(cid:4)hrt zu weniger Mikrorissen
β
des Materials im Endlager. Au erdem k(cid:4)onnen Zirkoniumkeramiken Mischkristalle mit
Aktiniden bilden. Um stabile Mischkristalle mit kubischer Symmetrie zu erhalten, wurde
in dieser Arbeit Yttriumoxid als Stabilisator f(cid:252)r die Zirkoniumkeramiken eingesetzt. Zur
Simulierung der radioaktiven Abf(cid:4)alle wurde Ceroxid und Neodymoxid gew(cid:4)ahlt.
Fu(cid:4)r die Synthese der Pulver wurde eine einfache Herstellungsmethode gew(cid:4)ahlt. Die
0−25
kubische Yttrium-stabilisierte Zirkonium-Matrix mit unterschiedlichen Cer- ( Gew
0−100
.%) und Neodym- ( Gew.%) Gehalt wurde mit der Coprezipitations-Methode syn-
110oC
thetisiert. Eine schonende Attritionsmethode wurde verwendet um die Pulver bei
600oC
zu trocknen oder um sie bei zu kalzinieren. Um eine hohe aktive Ober(cid:29)(cid:4)ache zu
erhalten, wurden die Pulver unter Azeton oder Propanol gemahlen. Fu(cid:4)r die Charakter-
isierung der Pulver wurden folgende Methoden angewendet: optische Mikroskopie und
β XRD
Granulometrie zur Bestimmung der Partikelgro(cid:4) e und -verteilung, zur Bestim-
TG/DSC
mung von strukturellen Eigenschaften, zum Nachweis von Phasenu(cid:4)berg(cid:4)angen
BET
und -Analyse zur Bestimmung der spezi(cid:28)schen Ober(cid:29)(cid:4)ache der Pulver. Nach diesen
grundlegenden Untersuchungen der Pulver wurden Tabletten mit unterschiedlichem Cer-
229MPa−1019MPa
und Neodym-Gehalt mit einem isostatischem Druck von gepresst.
1600oC
Diese wurden dann bei einer Temperatur von gesintert. Vor den Tests zur chemis-
chen Best(cid:4)andigkeit wurden unterschiedliche physikalische und mechanische Eigenschaften
der Cer- und Neodym-Keramik-Pellets bestimmt.
Die chemische Best(cid:4)andigkeit der Endprodukte wurde mit statischen und dynamischen
MCC − 1 MCC − 5
Auslaugungstests (entsprechend und ) mit unterschiedlichen Lau-
pH
gungsreagenzien bei relevanten -Werten untersucht. Der Ein(cid:29)uss der unterschiedlichen
Herstellungsprozesse auf das Auslaugungsverhalten der Produkte wurde untersucht und
bewertet. Die Resultate dieser chemischen Stabilit(cid:4)atstests zeigten, dass unterschiedliche
Strukturen einen Ein(cid:29)uss auf die Stabilit(cid:4)at der keramischen Materialien haben.
5
Abstract
Glass can undergo devitri(cid:28)cation with time and the incorporation of actinides with
α MeV α 106
their -decayin -rangeduetocompacted -recoiledirradiationover yearsstorage
leads to expansion and phase transformation of nuclear waste glasses. Moreover, the
maximumtolerableamountofactinidesthatcanbeincorporatedinglassisapproximately
5%. The development of ceramics adapted for nuclear-waste immobilization is the focus
of this dissertation.
(ZrY)O
The yttria-stabilized zirconia ceramic x is a promising candidate material for
both inert matrix fuel and waste form due to its high solubility of actinides, high chemi-
cal durability and exceptionally high stability under irradiation over a vast temperature
range. Cubic materials swell isotropically when damaged by radiation under di(cid:27)erent
strains. This leads to less microcracking in a repository. They are also capable of form-
ing isomorphous solid solutions with actinides. In order to obtain fully stabilized solid
solution with cubic symmetry, yttria oxide was chosen as a stabilizer for zirconia ceramics
and ceria oxide and neodymia oxide were chosen as stimulant wastes in the current study.
A simple fabrication method was chosen for the synthesis of powders. The cubic
yttria-stabilized zirconia matrix with di(cid:27)erent ceria contents (0-25 wt %) and neodymia
(0-100 wt %) contents was synthesized using the co-precipitation method. The mild
110oC 600oC
attrition method was used to dry powders at or calcine powders at . Both
acetone and propanol milling solvent media were used to get active (cid:28)ne powders for
neodymia ceramics. Then, the di(cid:27)erent characterizations of powders were performed
using the following methods: agglomerate size and distributions of powder by optical
microscopy and granulometry, properties of the powder and its structural appearance
XRD
by , phase transition and structural formation of the powder with a temperature
TG/DSC BET
gradient by , surface area and crystalline size of the powder by .
Aftertheseinitialinvestigationsintothecharacteristicsofthepowders,pelletswithdif-
229MPa−
ferent ceria and neodymia contents, were compacted using isostatic pressure of
1019MPa 1,600oC
. Then, they were sintered at a temperature of . Some physical and
mechanical properties of ceria ceramic and neodymia ceramic pellets were investigated
before chemical durability tests, were carried out.
Thechemicaldurabilityofthe(cid:28)nalproductswassubsequentlyinvestigatedusingstatic
MCC−1 MCC−5
and dynamic leaching tests, according to and with di(cid:27)erent leachants
pHs
at interesting . The in(cid:29)uence of the di(cid:27)erent fabrication processes on the leachability
of fabricated products was investigated and evaluated. The results from these chemical
durability tests were then used to determine the e(cid:27)ect that di(cid:27)erent structures have a
stability of the ceramic materials.
6
1 Introduction
1.1 Role of nuclear energy
Nuclear energy provides a signi(cid:28)cant contribution to the overall energy supply worldwide.
GWe
With 145 operating reactors, producing an overall power out put of 25 , the resulting
TWh
energy generation of 850 per year provides 35 % of the total electrical energy in the
European Union. Used fuel discharged from nuclear power plants constitutes the majority
of nuclear waste in countries that do not undertake reprocessing.
Fresh nuclear fuel essentially consists of uranium dioxides and heat generated from
the (cid:28)ssion of uranium atoms in the reactor. In addition to (cid:28)ssion products, generation
of plutonium and minor actinides result. During the reactor operation, (cid:28)ssion products
build up to such a level and reduce the mechanical integrity of the fuel. At this stage,
spent fuel has to be removed from the reactors and be replaced by new fresh fuel. Most
of the initial radioactivity in spent fuel is caused by (cid:28)ssion products and other short-life
radionuclides generated by the activation of metallic cladding material.
1.2 Current management of nuclear spent fuel cycles
1.2.1 Actinides, (cid:28)ssion products and radiotoxicity
FPs
Spent nuclear fuel consist of 96 % uranium, 1 % plutonium and 3 % (cid:28)ssion products
MAs Mt Mt
and "minor actinides" ( ) [1]. Each year 70 to 80 of new plutonium is left
SNF
in spent nuclear fuel ( ) globally [1]. In addition to plutonium and uranium, "minor
237Np 241Am
actinides", (with a half-life 2.1 million years), (with a half-life of 433 years)
234Am 244Cm FPs
+ and (with a half-live of 18.1 years) and (cid:28)ssion products ( ) such as
99Tc 129I 1.57 × 107 137Cs
(with a half-life of 213000 years), (with a half-life of years),
90Sr
(with a half-life of 30 years), (with a half-life of 29.1 years ) are generated in reactors.
During 50 years of nuclear energy production, approximately 1,100 metric tons of plu-
Np,Am,Cm
tonium and substantial quantities of minor actinides, such as generated in
Mt
nuclear fuel cycle. Additionally, 300 are held in weapons programs. Commercially
Mt
generated spent nuclear fuel accounts for 200 of plutonium. This plutonium is distin-
Pu U
guished for fabrication in a mixed-oxide fuel ( and ) mainly in UK and France. Over
Mt
900 remain in spent fuel which is stored in 236 nuclear power plants in 36 di(cid:27)erent
239Pu
countries. Reactor-grade plutonium (60 % ) could potentially be used for the de-
10kg 239Pu
veloping of nuclear weapons and nuclear devices can be made with less than .
The disarmment process will produce more weapons-grade plutonium, as well as hundred
HEU
of tons of highly enriched uranium ( ) which will be burnt in commercial reactors.
137Np 241Am 234Am 244Cm
The current global production rates of , + , and , are 3.4 , 2.7
Mt
and 0.35 per year respectively. These nuclides are important potential contributors
to the calculated exposure limits for humans over the long periods envisaged for their
geological disposal.
As Figure 1 [2] illustrates, it is evident that the radiotoxicity potential is dominated
241Am 243Am 239Pu 240Pu 242Pu
by the elements, americium ( , ) and plutonium ( , , ). After
Pu
approximately 1,000 years, the primary Am isotopes have mostly decayed while the
isotopes dominate. These will be substituted after 200.000 years by neptunium. The
99Tc 129I
relatively long-lived (cid:28)ssion products, and , have a radiotoxicity potential of one
7
Np
order of magnitude below . The range of radiotoxicity levels that will be accepted
by the public and by politicians is currently being discussed. The results will in(cid:29)uence
research concerning the necessity and degree of separation.
The release of actinides from spent fuel or other actinide-bearing solids is a critical
aspect of the evaluation of the long-term performance of geologic repositories. The long-
lived radiotoxicity potential introduced into a repository is basically determined by the
U Pu
losses of the actinides + during reprocessing, if there is no direct disposal.
1E+10
DPWWRR-U U-Z-Fyukelul s
Pu-chain
Pu-Kette 3355 GGWWdd//tt HSMM
1E+09
GWaWa el 1E+08 AmAm-c-hKaeintt e
G
v /v/
SS CCmm-c-Khaeitnte
al in ntial / 1E+07
tie AAccttiinniiddeens
nt
eo
tp
os
pg 1E+06
n
yu
citrd NNpp-c-hKaeitnte
ih
xä
of
Te 1E+05 Tc-99
G
UU--Kcehtatien
1E+04 strSutcrtuukratul rmmaatteerriaialsl FSispsaioltnpprroodduukcttes
I-129
1E+03
10 100 1.000 10.000 100.000 1.000.000 10.000.000
TZimeite / /J yahearers

Figure 1: Time-dependent toxicity potential (ingestion) of decay chains by elements of
PWR
spent fuel [2]
1.2.2 Partitioning-transmutation for actinides
The successful management of actinides has an important impact on the development
strategy for advanced nuclear fuel cycles, weapons proliferation, and geologic disposal
of high-level radioactive wastes. The minor actinides, although present in relatively low
concentrationsifdisposed, arehazardoustolifeformswhenreleasedintotheenvironment.
8
As a results, the current strategy consists of the separation of long-lived actinides, and
the (cid:28)ssion products from the biosphere. This avoids possible long-term harmful e(cid:27)ects.
Their disposal requires isolation from the biosphere in stable, deep geological formations
for long periods of time (some hundred thousand years) until their radioactivity decreases
naturally through the process of radioactive decay.
Today, it is di(cid:30)cult to accept nuclear waste disposal because the processes of separat-
ing this waste from the biosphere in order to avoid possible harmful e(cid:27)ects is a lengthy
one that structures into the future, implicating generations yet to come. To address this
problem, scientists are looking for ways of signi(cid:28)cantly reducing both the volume and ra-
diotoxicity of the waste by shortening the lengthy times required for the separate disposal
of this waste.
This is the aim of the partitioning and transmutation (P&T) activities worldwide.
Various international concepts on how these minor actinides could be separated from the
waste(partitioning)andconvertedintoshorterlivedlessharmfulnuclides(transmutation)
are being investigated. If it were possible to separate these minor actinides completely
from spent fuel and use nuclear reactors, or another facility to convert them into stable, or
shorter-lived nuclides, the disposal and management of wastes would greatly facilitated,
and public acceptance of nuclear energy could thus be raised.
Partitioning refers to aqueous or pyro-processing methods, which are used to separate
(partition)thevariouscomponentsofusedfuel: uranium, plutonium, minoractinides, and
volume (cid:28)ssion products. Today, the current technology used to reprocess spent fuel that
PUREX
has been removed from a reactor, is based on the aqueous process, in which the
spent fuel is dissolved in nitric acid. Using an organic solvent, plutonium and uranium are
recovered and used for fresh fuel fabrication. Thereafter, the aqueous ra(cid:30)nate, containing
minor actinides and (cid:28)ssion products, is vitri(cid:28)ed for disposal. The extraction technique
can also be extended to the extraction of neptunium. However, americium and curium
cannot be directly separated in this process.
PUREX
The recovery of neptunium can be achieved e(cid:27)ectively using a modi(cid:28)ed
VI Np U Pu
process: by adjusting the oxidation state to ( ), is extracted with and and
Np
can be separated from the heavy metal stream, either as single element fraction, or co-
extracted with uranium, or routed towards the plutonium stream. Some long-lived (cid:28)ssion
129I 99Tc PUREX
products such as and could also be separated within an improved
process.
The long-term radiotoxicity level can only be reduced by approximately one order
U Pu
of magnitude by the separation of and using the current reprocessing technique
U Pu Np Cm Am
(Figure.2). If and are separated at a higher level and if , and are ad-
HLLW
ditionally separated from the high-level liquid waste( ), the radiotoxicity potential
HLLW
level of will be lowered by about four orders of magnitude after approximately
1,000 years to the level of fresh nuclear fuel. This can be seen in Figure.2 [2].
U
The success of P&T depends on the partitioning quantity of and Pu (so far 99.8 %)
Np Am Cm
as well as on an additional e(cid:27)ective partitioning of , and from the high-level
HLLW
radioactive waste ( ).
An/Ln
However, fractions arise in all of the above-mentioned processes and further
Am Cm
partitioning causes considerable problems. The present oxidation state of and in
nuclear fuel dissolution liquors is (III). The oxidation state of lanthanide (cid:28)ssion products,
which represent about one third of the total mass of the (cid:28)ssion product inventory in the
9