Citation
Nuclear waste glass leaching in a simulated granite repository

Material Information

Title:
Nuclear waste glass leaching in a simulated granite repository
Creator:
Zhu, BingFu, 1946-
Publication Date:
Language:
English
Physical Description:
xvi, 212 leaves : ill. ; 28 cm.

Subjects

Subjects / Keywords:
Alkalies ( jstor )
Bentonite ( jstor )
Boron ( jstor )
Corrosion ( jstor )
Granite ( jstor )
Groundwater ( jstor )
Ions ( jstor )
Leaching ( jstor )
Nuclear waste ( jstor )
pH ( jstor )
Dissertations, Academic -- Materials Science and Engineering -- UF
Leaching ( lcsh )
Materials Science and Engineering thesis Ph. D
Radioactive waste disposal in the ground ( lcsh )
Genre:
bibliography ( marcgt )
non-fiction ( marcgt )

Notes

Thesis:
Thesis(Ph. D.)--University of Florida, 1987.
Bibliography:
Bibliography: leaves 203-211.
General Note:
Typescript.
General Note:
Vita.
Statement of Responsibility:
by BingFu Zhu.

Record Information

Source Institution:
University of Florida
Holding Location:
University of Florida
Rights Management:
Copyright [name of dissertation author]. Permission granted to the University of Florida to digitize, archive and distribute this item for non-profit research and educational purposes. Any reuse of this item in excess of fair use or other copyright exemptions requires permission of the copyright holder.
Resource Identifier:
000948893 ( ALEPH )
16904849 ( OCLC )
AER1000 ( NOTIS )
AA00004851_00001 ( sobekcm )

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Full Text
I;


NUCLEAR WASTE GLASS LEACHING
IN A SIMULATED GRANITE REPOSITORY
BY
BINGFU ZHU
A DISSERTATION PRESENTED TO THE GRADUATE SCHOOL
OF THE UNIVERSITY OF FLORIDA IN
PARTIAL FULFILLMENT OF THE REQUIREMENTS
FOR THE DEGREE OF DOCTOR OF PHILOSOPHY
UNIVERSITY OF FLORIDA


To my mother and late father


ACKNOWLEDGMENTS
The author acknowledges his gratitude to Dr. David E. Clark for
guidance and encouragement throughout his research. He also is
greatly indebted to Drs. Larry L. Hench, Lars Werme and George G.
Wicks for their advice and encouragement. The author thanks Drs.
Alexander Lodding, Christopher D. Batich, Stanley R. Bates and Gar B.
Hoflund for their timely suggestions, helpful discussions and review
of this dissertation.
The author wishes to thank Dr. Cheng Jijian for introducing him
to the field of chemical durability of glasses. Without his
guidance, the fulfillment of this research could not be possible. He
also thanks his wife, Jisi, for her support and encouragement.
iii


TABLE OF CONTENTS
Page
ACKNOWLEDGMENTS iii
LIST OF TABLES vi
LIST OF FIGURES viii
ABSTRACT xv
CHAPTERS
I INTRODUCTION 1
II PREVIOUS WORK ON NUCLEAR WASTE GLASS LEACHING 13
Laboratory Studies 13
General Considerations 13
Effect of Flow Rate 17
Surface Film Formation 20
Molecular Mechanism of Aqueous Dissolution 25
Systems Interaction Tests 29
Burial Studies 33
III RESEARCH OBJECTIVES, APPROACH AND SUMMARY OF
CONCLUSIONS 36
Research Objectives and Approach 36
Major Conclusions 37
IV MATERIALS AND METHODS 39
Glass Compositions and Characterization ...39
Burial Samples 39
Glass Quality 44
Laboratory Samples 48
Stripa Field Tests 48
Sample Assemblies, Minicans and Pineapple Slices...48
Stripa Repository 51
Burial and Retrieval 55
Disadvantage of the Burial Test Method 60
Similar Tests Being Used in MIIT Studies at WIPP...61
Laboratory Tests of Simulated Corrosion 62
IV


Analytical Techniques 67
Solid State Analyses 67
Solution Analyses 85
VTEST RESULTS 87
Field Test Results 87
General Observation 87
Results with ABS Glasses 89
Results with SRL Glasses 116
Effect of Glass Heterogeneities 125
Laboratory Test Results 135
Modified MCC-1 Static Leach Tests 135
Single-Pass Flow Tests and Static Tests Using
Rock Cups 140
VIDISCUSSION 152
ABS Glasses 152
SRL Glasses 1 61
A Model of Alkali Borosilicate Glass Leaching 168
Effect of Glass Composition 179
Influence of Repository Variables 185
Ground Water Chemistry 185
Effects of Repository Materials 1 88
Effect of Temperature 193
Comparison of Field and Laboratory Test Results 193
VIISUMMARY 197
REFERENCES 203
BIOGRAPHICAL SKETCH 212
v


LIST OF TABLES
Table Page
1-1 Quantity and Radioactivity of High-Level Nuclear Wastes
in Some Developed Countries 5
1-2 Candidate Waste Forms Considered for Geologic Disposal
of High-Level Waste 3
4-1 Nominal Waste Glass Compositions (wt) Used in the
Stripa Burial 40
4-2 Sample Matrix of the Stripa Burial Tests 45
4-3 Variations in Spectral Characteristics of SRL Waste
Glasses 46
4-4 Nominal Composition of Black Frit 165-Mobay Glass 49
4-5 Average Major/Minor Chemical and Mineral Constituents in
Stripa Granite 54
4-6 Ground Water Composition and pH Measured in this Study
within the 1-month Test Hole at Stripa. Concentration
mg/L 58
4-7 Ground Water Composition for the Stripa Granite,
Literature Values 59
4-8 Sample Matrix of the Laboratory Tests 68
4-9 Characteristics of Analytical Techniques 70
5-1 Composition of Ground Water Collected from the Boreholes
where SRL Glass Pineapple Slice Assemblies Had Been
Buried 91
5-2 Gram*Atoms of Elements Remaining at Gel Mid-Plateau
and Outer Region of the Altered Glass Surface Based
on 100 GramAtoms of Unleached ABS 118 Glass after
12-Month, 90C Burial in Stripa 119
vi


5-3 Relative Concentrations (Ratio to Si) at the Black Frit
165-Mobay Glass Surface after Static Leaching in the
Rock Cup Test. Data Are from EDS Analysis 147
6-1 90C Glass Leach Rates During 12- to 31-Month Period
( pm/year) 155
6-2 SIMS Compositional Analysis of Glass/Glass, Glass/
Bentonite and Glass/Granite Interfaces for ABS 39,
ABS 41 and ABS 118 after 12-Month, 90C Stripa Burial
(Gram*atoms Remaining Based on 100 Gram*atoms of
Unleached Glass) 158
6-3 Coordination Number and Bond Strength of Most Oxides in
Alkali Borosilicate Nuclear Waste Glasses 170
6-4 90C ABS Glass Leach Rates During 7-12 Month Period 191
7-1 Estimated Boron Depletion .Depths (pm) after 300 Years of
the Thermal Period of Storage for the Six Nuclear Waste
Glasses 201
7-2 Estimated Boron Depletion Depths (pm) after 10^
Years of Storage for the Glass/Glass Interfaces of SRL
Simulated Nuclear Waste Glasses 202
vi i


LIST OF FIGURES
Figure Page
1-1 Flow diagram showing the reprocessing of the spent
nuclear fuel 2
1-2 Schematic showing the glass waste form in a geological
repository 6
1-3 Glass structure containing dissolved wastes 11
2-1 Research activities on leaching of nuclear waste glass 1 I
2-2 Plot showing the total mass loss per unit area as a
function of flow rate 19
2-3 The five types of glass surfaces and six surface
conditions resulting from glass-environment interactions... 21
2-4 Ratio of normalized solubility to NLg^ (20 g/m^)
for CaC03, SrC03, Nd(0H)3, Fe(0H)3 and Zn(0H)2
in MCC-1 28-day test at 90C in solutions of different
pH 26
2-5 The Si leachability of a borosilicate glass immersed
in a 5-day static 23C solution buffered to various
pH values 27
2-6 Calculated Si concentrations in the surface layer and
bulk solution based on the surface layer diffusion
and pH as a function of leaching time. A diffusion
coefficient of 10 cm2/sec was assumed 30
4-1 Pineapple slices of glass, granite, stainless steel,
Ti, Pb and compacted bentonite before burial 43
4-2 Representative FT-IRRS spectra of SRL glass pineapple
slices before burial 47
4-3 A minican assembly 50
4-4 A typical pineapple slice assembly 52
viii


4-5 Seven preburial pineapple slice assemblies with different
sample stacking sequences for SRL simulated nuclear waste
glasses 53
4-6 Location within Stripa where SRL samples were buried 56
4-7 Diagram illustrating the position of the samples in the
Stripa mine during burial 57
4-8 Schematic of experimental configuration of static leach
test 63
4-9 A corrosion cell in the flowing test 65
4-10 Schematic of experimental configuration of continuous
flow test 66
4-11 Sampling depths with various techniques used in this
study 69
4-12 Light micrograph of SRL 131 + 29.8% TDS glass with
crystallites (100X) 72
4-13 Light micrograph of a typical glass surface after
polishing to 600 grit surface finish (100X) 73
4-14 FT-IRRS analysis of SRL 165 + 29.8% TDS glass/glass
interface prior to and after 2-year burial in Stripa 76
4-15 SEM micrograph of a typical glass surface after
polishing to 600 grit prior to leaching 77
4-16 EDS analysis of an uncorroded SRL 131 + 29.8% TDS glass
surface 78
4-17 SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 90C burial in Stripa. The atomic
concentrations of all cations (except H) are summed up
to 100$ 80
4-18 SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 90C burial in Stripa. Data are presented
as gram*atoms of various cations remaining in the leach
layer at certain depth based on 100 gram*atoms of
unleached glass 83
5-1 A typical assembly after burial in Stripa mine.
Bentonite coating can be observed on the outer surface
due to bentonite swelling 88
ix


5-2 Schematic of glass/glass interface illustrating several
types of surface areas resulting from water and/or
bentonite intrusion 90
5~3 FT-IRRS spectra of glass ABS 39 (a) and ABS 41 (b) before
and after 31-month, 90C Stripa burial 92
5-4 SIMS depth profiles for (a) ABS 39 (Al-corrected) and
(b) ABS 41 (Si-corrected) after 31-month, 90C Stripa
burial 94
5-5 Light micrographs (100X) of glass 39 (a) glass/glass,
(b) glass/granite and (c) glass/bentonite interfaces and
glass ABS 41 (d) glass/glass, (e) glass/granite and
(f) glass/bentonite interfaces after 31-month, 90C
Stripa burial 97
5-6 SIMS depth profiles of boron for glass ABS 39 (a)
and glass ABS 41 (b) after 31-month, 90C Stripa burial....99
5-7 FT-IRRS spectra of glass/glass, glass/granite and glass/
bentonite interfaces for nuclear waste glass ABS 118
buried in Stripa at 90C for (a) 2 months and (b) 12
months. Also shown is the spectrum of a preburial
glass surface 100
5-8 Light micrographs (100X) of glass ABS 118 after
2-month burial, (a) glass/glass, (b) glass granite,
and (c) glass/bentonite interfaces, and after 12-month
burial, (d) glass/glass, (e) glass/granite, and (f)
glass/bentonite interfaces at 90C in Stripa 102
5-9 SIMS depth compositional profiles of (a) B; (b) Cs,
Sr; and (c) Fe, U for ABS 118 glass/glass interface
after 2- and 12-month, 90C burial in Stripa. The data
have been corrected using A1 concentration 103
5-10 SIMS depth compositional profiles of (a) Si, H, Na,
Li, K; (b) LD (including La, Ce, Pr, Nd and Y), P, Sn;
(c) Ca, Zn, Ba; and (d) Zr, Mo, Ni, Cr, Si for ABS 118
glass/glass interface after 12-month, 90C burial in
Stripa 105
5-11 FT-IRRS analysis of ABS 118 glass/Pb, glass/Ti and
glass/Cu interfaces after (a) 2-month and (b) 12-month
90C burial in Stripa 108
x


5-12 Light micrographs of ABS 118 glass surfaces after
90C Stripa burial for 2 months, (a) glass/Pb, (b)
glass/Cu, and (c) glass/Ti interfaces, and for 12
months, (d) glass/Pb, (e) glass/Cu, and (f) glass/Ti
interfaces 109
5-13 Boron profiles of ABS 118 (a) glass/Pb, (b) glass/Cu
and (c) glass/Ti interfaces after 2- and 12-month, 90C
burial at Stripa 111
5-14 Cs and Sr profiles of ABS 118 (a) glass/Pb, (b) glass/Cu
and (c) glass/Ti interfaces after 2- and 12-month, 90C
burial at Stripa 112
5-15 Fe and U profiles of ABS 118 (a) glass/Pb, (b) glass/Cu
and (c) glass/Ti interfaces after 2- and 12-month, 90C
burial at Stripa 113
5-16 SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90C Stripa
burial, Si, H, Li, Na and K profiles 114
5-17 SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90C Stripa
burial, Ca, Zn and Ba profiles 115
5-18 SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90C Stripa
burial, LD, Pb, Cu and Ti profiles. LD stands for the
sum of La, Ce, Pr, Nd and Y 117
5-19 SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90C Stripa
burial, Zr, Mo, Ni and Cr profiles 118
5-20 RBS analysis of ABS 118 glass/Pb, glass/Ti and
glass/Cu interfaces after 12-month, 90C burial in
Stripa 120
5-21 FT-IRRS analysis of the glass/glass interface for three
SRL glasses after 2 years of burial in Stripa 121
5-22 SEM micrographs of glass surfaces in contact with glass
of the same composition during 2-year burial at 90C
in Stripa: (a) SRL 131 + 29.8? TDS, (b) SRL 165 +
29.8? TDS, (c) SRL 131 + 35? TDS and (d) an uncorroded
glass surface 123
5-23 SIMS depth profiles of SRL glass surfaces after 2-year
Stripa burial at 90C 124
xi


5-24 X-ray diffraction pattern for powders prepared from
devitrified SRL 131 + 29.8% TDS glass 126
5-25 SEM-EDS analysis of preburial SRL 131 + 29.8% TDS
glass: (a) homogeneous glass surface and (b) partially
devitrified glass surface 127
5-26 SEM analysis of SRL 131 + 29.8% TDS glass surfaces in
contact with bentonite, Stripa burial at 90C: (a)
homogeneous glass, 1-month burial; (b) partially
devitrified glass, 1-month burial; (c) partially
devitrified glass, 3-month burial; and (d) partially
devitrified glass, 6-month burial 129
5-27 FT-IRRS analysis of SRL 131 + 29.8% TDS glass surfaces
in contact with bentonite, Stripa burial at 90C: (a)
homogeneous glass, 1-month burial; and partially
devitrified glass (b) 1-month burial; (c) 3-month
burial; and (d) 6-month burial 131
5-28 EDS analysis of SRL 131 +29.8? TDS glass surfaces in
contact with bentonite, Stripa burial at 90C: (a)
homogeneous glass, 1-month burial; and glass matrix of
partially devitrified glass, (b) 1-month burial; (c)
3-month burial; and (d) 6-month burial 133
5~29 EDS analysis of crystal areas of partially devitrified
SRL 131 + 29.8% TDS glass surfaces in contact with
bentonite, Stripa burial at 90C: (a) for 1 month,
(b) for 3 months and (c) for 6 months 134
5-30 FT-IRRS analysis of SRL 165 + 29.8 TDS glass before and
after leaching for 28 days at 90C in (a) deionized
water with SA/V = 0,1 cm-1, (b) Stripa ground water
with SA/V = 0.1 cm (c) Stripa ground water with
SA/V = 1.0 cm 1 and (d) saturated Stripa ground water
with SA/V = 0.1 cm 1. Also shown is a spectrum for
the glass/glass interface after 3~month Stripa burial 136
5-31 SIMS analysis of SRL 165 + 29.8% TDS waste, laboratory-
corroded, 1 month 90C in Stripa water with SA/V =
1 .0 cm'1 139
5-32 Solution pH vs time for Black Frit 165-Mobay glass
corroded under static and flow conditions 1 41
5-33 Concentrations of Si, B, A1 and Li in the single pass
flowing ground water at 0.3 ml/hr as a function of
leaching time for Black Frit 165-Mobay glass samples 142
xi 1


5-34 Normalized leach rates of Li, B and Si as a function
of time under flowing (at 0.3 ml/hr) conditions for
Black Frit 165-Mobay glass with SA/V = 1.0 cm-1.
Also shown are the weight losses for glass samples
leached under static and flow conditions with SA/V
= 1 .0 cm"1 144
5~35 EDS analysis of Black Frit 165-Mobay glass leached in
the rock cup test at 90C with SA/V = 1.0 cm 1 in
ground water under static conditions 145
5-36 FT-IRRS analysis of Black Frit 165-Mobay glass leached
in the granite rock cup test at 90C under static
conditions with SA/V = 1.0 cm 1 148
5-37 SIMS depth profiles for Frit 165-Mobay glass leached
in the granite rock cup tests at 90C with SA/V =
1.0 cm (a) under static conditions and (b)
under flow conditions (0.3 mL/hr) 150
6-1 Time dependence of reaction layer thickness for glass
ABS 39 (a) and ABS 41 (b) after 31_month, 90C Stripa
burial 153
6-2 Boron depletion depth vs burial time for the glass/
glass, glass/granite and glass/bentonite interfaces.
Three ABS glasses are compared 1 60
6-3 Penetration depth as a function of leaching time for the
SRL glasses either buried in contact with glass,
stainless steel, granite or bentonite in Stripa mine, or
leached in Stripa ground water with SA/V = 0.1 or 1.0
cm 1 in laboratory 162
6-4 SIMS compositional profiles of SRL 165 + 29.8? TDS
glass/bentonite interface after 24-month, 90C burial
in Stripa 165
6-5 Five modes of corrosion in partially devitrified
alkali borosilicate simulated nuclear waste glass:
(a) leaching of the glass matrix; (b) enhanced attack
of the glass-crystal interface; (c) pitting of the
polycrystalline phase at grain boundaries; (d) surface
films enriched in the less soluble multivalent
species; and (e) crystallite stripping 167
6-6 Stability of B2O2 and SO2 in aqueous solution
at 25C as a function of pH
xiii
173


6-7 Schematics showing (a) the altered alkali borosilicate
glass surface and the compositional profiles after
leaching based on the model proposed in this
dissertation and (b) the altered glass surface
based on Grambow's model 176
6-8 The density index curve for three SRL glasses after
2-year burial in Stripa at 90C 1 80
6-9 Compositional ternary diagram showing the direction of
increasing boron depletion depth. F^O represents
alkali metal oxide, represents A^O^ and
Fe2C>2, and WP stands for waste products 181
6-10 The boron depletion depth as a function of
(SiOg + A^O^/CF^O + 820^) wt ratio
in glasses 183
6-11 Schematic illustrating the relationship between
concentration, contact time and leach rate 186
6-12 The boron depletion depth as a function of burial time
for ABS 118 glass/glass, glass/granite, glass/Pb, glass/
Cu and glass/Ti interfaces after 90C Stripa burial 190
6-13 SIMS compositional profiles of SRL 165 + 29.8% TDS
glass/glass interface after 8-10C Stripa burial for
2 years 19^
xiv


Abstract of Dissertation Presented to the Graduate School
of the University of Florida in Partial Fulfillment of the
Requirements for the Degree of Doctor of Philosophy
NUCLEAR WASTE GLASS LEACHING
IN A SIMULATED GRANITE REPOSITORY
BY
BINGFU ZHU
May 1987
Chairman: Dr. David E. Clark
Cochairman: Dr. Larry L. Hench
Major Department: Materials Science and Engineering
Burial experiments of three Savannah River Laboratory (SRL) and
three Swedish alkali borosilicate (ABS) simulated nuclear waste
glasses were conducted to evaluate the resistance of these glasses to
ground water attack under repository-like conditions. Glass samples
were buried in the boreholes at a depth of about 350 meters below the
surface in the Stripa granite at either ambient mine temperature
(8-10C) or 90C. Included in the same boreholes were other
potential waste package components. Glasses were also leached in the
Stripa ground water contained in a leaching vessel under the
laboratory simulation conditions. The leached surfaces were
characterized using SEM-EDS, FT-IRRS, SIMS, RBS and optical
microscopy. Differences in glass leach rate were observed among the
six compositions with SRL 165 + 29.8% TDS being the lowest. Results
show that durabilities of the SRL composite nuclear waste glasses
XV


were increased by approximately six times when frit 131 was
substituted by frit 165. An increase of waste loading of SRL 131
glass from 29.8 wt% to 35 wt% decreases the teachability by a factor
of 2.
The leach rates of buried samples based on boron extraction at
90C ranged from 0.3~3 pm/year for the glass/glass interfaces of all
glass formulations. These values are at least two orders of
magnitude lower than those for glasses leached using MCC-1 static
leaching procedures and deionized water. The Stripa repository-like
conditions can be simulated in the laboratory using Stripa ground
water and high SA/V ratios (_> 1.0 cm-1). Comparison of the
laboratory test results with field test results indicates that the
leaching mechanisms were similar under these test conditions. One of
the advantages of the laboratory simulation testing is saving of time
since glass leaches faster under the laboratory-controlled conditions
than under field-leach environment.
A model, based on glass structure and thermodynamic
considerations, is proposed to describe alkali borosilicate glass
leaching under repository-like conditions.
xvi


CHAPTER I
INTRODUCTION
The increasing use of nuclear energy for electric power
generation and the expanding application of radioisotopes in various
fields are inevitably associated with the production of growing
amounts of nuclear wastes. These wastes, which result from
fabrication, use and reprocessing of nuclear fuels, contain a variety
of hazardous materials. Hence, their disposal must ensure a low
probability of human contact.
Major types of nuclear wastes include high-level (HLW),
transuranic (TRU), low-level (LLW), uranium mine and mill tailings,
decontamination and decommissioning wastes, and gaseous effluents.
High-level wastes are usually further divided into those resulting
from either weapons production (defense waste) or commercial power
reactors.
Upon removal from the nuclear reactors, the depleted fuel is
stored under water for several months to permit the short-lived
fission products to decay. One of the options is to send the fuel
pellets to a chemical-reprocessing plant to recover the uranium and
plutonium, which are then available to make new fuel [1]. As shown
in Fig. 1-1, the reprocessing generally consists of dismantling
reactor fuel in a manner that permits dissolution of the core
material of the nuclear fuel pellets without dissolving their
1


Spent
fuel
clad
material
dismantling
->
core
material
solvent extraction
or ion exchange
U and Pu
recovered
Waste
Fig
1 -1 .
Flow diagram showing the reprocessing of the spent nuclear fuel [1].


3
corrosion resistant cladding [1], The resulting solution is
subsequently treated by several cycles of solvent extraction or ion
exchange to recover, separate and purify the residual uranium and
plutonium. At Savannah River Plant, Aiken, South Carolina; in
Hanford Reservation, Richland, Washington; and in Idaho National
Engineering Laboratory, outskirts of Idaho Falls, Idaho, there are
large facilities owned and operated by the United States government
that reprocess spent fuel coming out of the reactors used for making
weapons. However, only one commercial reprocessing plant, at West
Valley, New York, was ever operated in the U.S. Currently there is
no reprocessing of spent fuel coming out of commercial reactors in
the United States.
High-level nuclear wastes, whether reprocessed or not, contain
virtually all of the nonvolatile fission products, small amounts of
uranium and plutonium and all the other actinides formed by
transmutation of the uranium and plutonium in the reactors.
They can be generally characterized by their very intense,
penetrating radiation and their high heat-generation rates. The
fission products and actinides are the major concern since they
undergo spontaneous decay and emit radioactivity in the form of a and
1 37 90
8 particles, and Y-rays. The two elements, Cs and Sr are of
most concern due to the relatively high concentration in the waste
and their decay time (i.e., 30 years)[2] and concern for their
90
incorporation in body tissues, especially Sr in bones. When
decaying, they give off both heat and radioactivity for about 700
years [2]. The actinides including U also emit radioactivity and


4
? vq
heat during decayfor example, about 25,000 years for Pu J the
most abundant transuranium actinide [2].
Table 1-1 lists the quantity and radioactivity of high-level
nuclear wastes in some developed countries [3_5]. There are over
5 2
3X10" m" defense HLW stored at three government sites in the United
States. These defense wastes contain 1.6X10^ Curies of radioactivity
(1 Curie = 3-7X1010 disintegrations per second)[3]. There are over
2X1 O'* m^ commercial HLW containing about 1.1X101^ Curies of
radioactivity in the form of spent fuel in the United States [3].
The total amount of HLW in Europe, Japan, the United States and
U.S.S.R. is estimated to be 10.2X10' containing about 2.9X101^
Curies.
One method for disposal of HLW is immobilization in a high-
integrity solid waste form followed by emplacement in a mined cavern
at a suitable geologic repository [6]. As shown in Fig. 1-2, this
disposal system relies on multiple barriers to prevent the release of
radionuclides. The system includes
(1) solid waste form, a combination of host material (glass in
the illustrated case) and waste. The waste is incorporated
homogeneously in the host material to reduce the risk for
dispersion.
(2) a metal canister such as stainless steel, which is welded to
form a hermetically sealed container after the waste form is
placed in it.
(3) a metallic overpack, of such materials as e.g., mild steel,
ductile iron, pure titanium, or titanium alloy (Ti Code-12),


5
Table 1-1. Quantity and Radioactivity of High-Level Nuclear Wastes
in Some Developed Countries.
Radioactivity Quantity
Form (Ci) (nr) Source
U.S. Defense
slurry
sludge
1.6X109
3X105
[3]
U.S. Commercial
spent
fuel,
sludge (West
Valley, NY)
1 .1 XI010
2X1 05
[3,4]
Europe Commercial
spent
fuel
-
2X1 05
estimated
Japan Commercial
spent
fuel
-
7X1 04
estimated
U.S.S.R.
slurry
sludge
-
2.5X10^
estimated
Total
2.9X1010
10.2X105
[5 ]


6
m
Fig. 1-2. Schematic showing the glass waste form in a geological
repository.


7
and nickel alloys [7], which serves as an additional barrier
for radionuclide containment.
(4) a sleeve, when required, which is used to assure clearance
for the retrievable package to facilitate its removal during
the retrieval period. It provides structural support
against geologic pressure forces and may also serve as a
barrier for radionuclide containment.
(5) backfill, the material contained between the other
engineered waste package components and the host rock, which
serves to facilitate heat transfer, load transfer and
compatibility of the other engineered waste package
components with the host rock. It may also serve as one of
the barriers for radionuclide containment and a sorptive
medium for radionuclide release. Swelling clays such as
bentonite, alone or in a mixture with quartz or other
minerals, are being considered as backfill materials.
(6) a buffer, the material used to facilitate conditioning of
the ground water, immobilization of radionuclides and
compatibility of materials.
(7) a filler, which is any material used to fill space between
other components of the engineered waste package and may or
may not have other specified functions.
Five years ago, there were seven candidate waste forms chosen
for geologic disposal of HLW in the United States (Table 1-2). After
a multifaceted assessment [8-11], borosilicate glass and Synroc (a
titanate-based polyphase crystalline ceramic material) were selected


8
Table 1-2. Candidate Waste Forms Considered for Geologic Disposal of
High Level Waste [8].
Waste Form
Comments
Borosilicate Glass
Primary Waste Form, U.S. Reference Waste
Form
Synroc-C,D
Alternative U.S. Waste Form
Tailored Ceramic
Semi-finalist U.S. Alternative Waste Form
High-Silica Glass
Semi-finalist U.S. Alternative Waste Form
FUETAP Concrete
Semi-finalist U.S. Alternative Waste Form
Coated Sol-Gel Particles
Semi-finalist U.S. Alternative Waste Form
Glass Marbles in Lead
Matrix
Semi-finalist U.S. Alternative Waste Form


9
from the seven as the primary waste form and first alternative,
respectively. The focus of this work is on borosilicate nuclear
waste glass.
There are two major reasons why glass was selected as the
primary waste form. First, any material used for encapsulating
radioactive wastes must be capable of surviving for at least 10,000
years in a wide range of severe environments. Glasses can meet this
requirement. The existence of natural glasses, such as obsidians,
basalts, or tektites, which are millions of years old, demonstrates
that glass can be formulated which will survive long-term
environmental exposures. Similarly, synthetic glasses of known
longevity or performance, such as Roman glasses buried in the
Mediterranean or exposed to ground water for nearly 2,000 years, also
demonstrate the potential long-term performance of nuclear waste
glass. Second, the process for producing nuclear waste glass is
fairly simple. It involves feeding a slurry of waste sludge and
glass frit to a continuous glass melter, from which the molten waste
glass is poured into a canister. Such a simple fabrication method
makes the remote control of the whole process possible, as
demonstrated in the United States and France in full-scale operations
[2,12].
In contrast, consolidation and synthesis of the mineral phases
in synroc require hot isostatic processing or uniaxial hot
processing, which complicates the remote production processes.
Although the uranium leach rates are higher and the waste loading is
lower for the glass form than for the crystalline ceramics,


10
borosilicate glass is currently the choice of most countries as the
primary waste form due to simplicity of fabrication, moderate waste
loading, intermediate product performance and radiation stability.
The list of candidate sites for the first repository in the
United States has been narrowed to three locationsone in Nevada in
volcanic tuff, one in Texas in salt, and one in Washington state in
basalt [2]. Other rock formations such as granite in Sweden have
been considered outside the United States [13]. A final decision on
the site in the United States is still several years away and will
require extensive testing and risk assessment.
The major concern when the waste is buried deep in the ground is
that it might come into contact with water and be transported back to
the earth's surface. Therefore, the resistance of the solid waste
form to underground water attack is a problem of major concern,
because the second innermost barrier (canister materials) is only
expected to survive about 1,000 years in a geologic environment [14].
A nuclear waste glass is defined as a single phase amorphous
material in which quantities of both radioactive and nonradioactive
oxides are dissolved. The concept of using glass as a host for
radioactive waste is based upon the radionuclides entering into and
becoming part of the random three-dimensional glass network. Figure
1-3 schematically illustrates a portion of an alkali borosilicate
glass network containing various radionuclides as constituents. The
4-
structural network of the glass is provided primarily by [SiO^] ,
5- 3-
[BOjj] and [BO^] polyhedra. Neighboring polyhedra are bonded
together by sharing strong ionic-covalent bridging oxygen bonds.


11
9
OXYGEN
S1UCON
BORON
Fig. 1-3. Glass structure containing dissolved wastes (adapted from
[15]).


12
+ 2 + 2
Other multivalent species such as Fe rare earths or actinides
are also generally bonded within the network by bridging oxygen
bonds. Low valence ions, such as Na+, Cs+, Sr+2, etc., are bonded
into the network by sharing various nonbridging oxygen bonds,
depending upon size and valence of the ions. This difference in type
of bonding in the glass network is responsible for the complex leach
behavior of nuclear waste glasses.
To date, there are only few data available regarding the
leaching behavior of nuclear waste glasses in the presence of a
variety of disposal system components [16-23]. In order to test
possible synergistic interactions of the materials in a nuclear waste
disposal system under repository-like conditions, in situ burial
experiments were designed. Such experiments approximated the
physical conditions of the repository more closely than laboratory
tests. Laboratory systems tests were also designed, when necessary,
to evaluate the effects of individual system variables on glass
leaching performance.
The primary objective of this dissertation was to determine the
leaching performance of the glass containing high-level nuclear
wastes* under a simulated repository condition and to investigate how
this is affected by the presence of other waste package components
and geologic conditions. In the process of achieving this goal a new
model of glass leaching was developed that satisfactorily describes
the observed results from both laboratory and field studies.
* The wastes used in this dissertation were simulated. It is assumed
that isotopes of the same element have similar chemical behavior.


CHAPTER II
PREVIOUS WORK ON NUCLEAR WASTE GLASS LEACHING
Extensive laboratory tests and some field tests have been
conducted using various combinations of reference materials in order
to evaluate their effects on glass leaching. Figure 2-1 summarizes
the research activities on nuclear waste glass leaching
performances. The laboratory tests performed include static and flow
experiments. In most of the laboratory tests, deionized waster was
used. Glasses were also leached in synthetic ground water, such as
silicate water and brine, and/or in the presence of other waste
package components. Relatively limited burial tests include a
15-year burial at Chalk River, Canada [24,25], a 9-year burial in
England [26] and more recently an initiated burial study in Belgium
[22]. More extensive work has been carried out in the Stripa mine in
Sweden [16-20]. The major focus of this investigation is on the
Stripa burial and laboratory systems interactions. The Waste
Isolation Pilot Plant (WIPP) program was designed based on the
experience from the Stripa burial test. This is the first burial
test to be conducted in the United States.
Laboratory Studies
General Considerations
For some time, the primary issue of concern regarding glass and
other HLW forms has been long-term stability in contact with hot
13


Fig
2-1
Research activities on leaching of nuclear waste glass.


15
repository ground waters in the event a canister is breached. In the
early 1980s, five tests were developed to determine the chemical
durability of waste forms [27,28] under either static (MCC-1P* and
MCC-2P) or flowing (MCC-4S and MCC-5S) leaching environments.
Maximum release by waste forms is determined using powders and
stirred solutions (MCC-3S). General acceptance of these test
methods, initiated by the Materials Characterization Center [27,28],
reduced inconsistency, improved communication and made possible the
comparison of data collected from different laboratories. This
facilitated the accumulation of an extensive data base on glass
leaching, including nuclear waste glasses.
In this paper, the term "leaching" is defined as release of
glass component oxides or elements through glass-aqueous solution
reactions without regard to mechanisms of release. The term
"corrosion" is also associated with deterioration of glass surfaces
due to the reactions that occur when water interacts with glass.
Therefore, these terms are used synonymously in this dissertation.
Most leach test data are reported for short periods of time,
i.e., 28 days or less. Such short-term data are frequently used to
compare the relative stability of waste forms and to study effects of
variables that control the rate of leaching. For example, Plodinec
et al. [29] used the approach of Newton and Paul [30] to predict
* Materials Characterization Center, Pacific Northwest Laboratory,
Richland, WA.


16
nuclear waste glass leaching based on thermodynamic aspects of its
chemical composition. They found a linear relationship between log
normalized mass loss of Si (g/m 28 days) and free energy of
hydration (kcal/mol) for a number of natural and synthetic glasses,
including simulated nuclear waste glasses. Comparing the corrosion
resistance of nuclear waste glasses to natural glasses and ancient
man-made glasses and/or relative thermodynamic stabilities allows
extrapolation of waste glass corrosion resistance to geologic times
[31,32].
Strachan [33] has reported results from a 1-year-leach test
using MCC-1 static test procedures. He found that a dramatic
decrease in the rate of leaching occurred after approximately 91
days. The PNL76-68* glass appeared to continue to alter, albeit at a
significantly reduced rate, even though the solution concentrations
of many elements were saturated or supersaturated with respect to
alteration phases. In his studies, glasses were leached in deionized
water, silicate water and brine at either 40, 70 or 90C with the
ratio of glass surface area to volume of leachout (SA/V) of 0.1
cm 1. Solid state analyses of the leached specimens indicated a
steady growth of two layers. The outer layer grown by precipitation
reactions on the original surface of the glass consisted
predominantly of zinc and silicon, thus indicating a zinc silicate
phase(s). An altered layer remained behind as the aqueous solution
* A nuclear waste glass composition developed by Pacific Northwest
Laboratories, Richland, WA.


17
leached soluble and moderately soluble material from the glass
matrix. Thus, this layer was rich in Fe, Nd, La, Ti, and depleted in
B, Cs, Na and Mo. The altered layer thicknesses for specimens
leached in deionized water, silicate water, and brine at 90C for up
to 1 year ranged from 30 to 50 pm.
The long-term data obtained by Bates et al. [3*0 using deionized
water and MCC-1 test conditions agree qualitatively with those
obtained by Strachan [33]. However, data of Bates et al. [34]
indicate that the normalized elemental mass losses for most elements
are constant after approximately 6 months of leaching, whereas
Strachan's data indicate that leaching losses are continuing for
several elements at the end of one year. The following equation
2
defines the normalized elemental mass loss NL.^ in g/m :
NL.
i
(2-1 )
where m^ = mass of element i in the leachate, g;
f^ = mass fraction of element i in the unleached specimen,
dimensionless;
p
SA = specimen surface area, m .
One important feature observed in results of Strachen [33] and Bates
et al. [34] is the preferential leaching of B and Na. The normalized
elemental mass losses for these two elements are larger than for Si.
Effect of Flow Rate
It is recognized that under certain conditions ground water will
flow through a geological repository and react with its contents.


18
Strachan et al. [35] have reported increased leach rates for Si and
Sr at a flow rate of 6 mL/h compared to static testing. Similar
results have been found by other workers [36]. Based on weighing the
samples before and after corrosion, the rate of leaching increased as
the flow rate was increased from 0.1 to 10 mL/h. Little difference
was observed between the static test and the test in which the flow
rate was 0.1 mL/h during the first 28 days.
It has been found [36] that at sufficiently low flow rates
(between 0.1 and 2 mL/h in Fig. 2-2) the total mass loss per unit
area of waste glass matrix components surface is directly
proportional to flow rate. The concentration of glass components in
the leachate is nearly independent of flow rate. In the case of some
matrix components (Si, Al), this concentration is determined by
saturation with respect to the surface of the glass, as modified by
the leaching process and possible alteration reactions. The modified
surface forms a barrier against migration.
At sufficiently high flow rates (> 2 mL/h in Fig. 2-2), the
release rate becomes constant, limited by the kinetics of the
leaching processes. In this case, corrosion products as well as
potential surface-passivating species are removed from the leaching
vessel, reducing the beneficial effects on both solution saturation
and protective surface film formation.
Present indications are that high flow conditions (>10 mL/h) are
very unlikely in a geologic repository [37]. Low flow conditions are
expected in the repository and the leach rate of the glass will be


19
!000?
CM
O'
LiJ
2:
IP
tx
Id
a.
%/)
co
o
CO
to
<
t
o
R-
SRL ¡31-29.3% TD3-3A
90 C, D.!. Water
SA/ V = 0.! cm-
Basalt ¡
tuff
granite Potential Flow Rates in Repositories
-l
suit
!
I
¡OF
o
o
J
G.i(mL/hr) C).5(mL/hr) i.Q(rnL/yr) ¡C(mLyS'ir)
4.6-m/yr) 23(m/yr) 46im/yr) 464(.n/yr)
LOG
LOW RATE
Figi 2-2. Plot showing the total mass loss per unit area as a
function of flow rate (adapted from [36]).


20
limited by the rate of transport of corrosion products from the
repository.
Surface Film Formation
Previous efforts to generalize the surface behavior of silicate
glasses proposed five types of glass surfaces and six surface
conditions to represent a broad range of glass-environment
interactions [38-40]. The type of surface is dependent on the
environmental history of the glass and may be defined in terms of
surface compositional profiles, as shown in Fig. 2-3. The ordinate
in Fig. 2~3 represents the relative concentration of SO2 (or oxides
in Type IIIB surface) in the glass and the abscissa corresponds to
the depth into the glass surface. If species are selectively
dissolved from the glass surface, the relative Si02 concentration
will increase producing a Si02-rich surface layer. If all species in
the glass are dissolved simultaneously (congruent dissolution), the
relative concentration of Si02 will remain the same as in the
original glass. When combinations of selective dissolution,
congruent dissolution and precipitation from solution occur, then any
one of the six surface conditions shown in Fig. 2-3 is possible.
Hench [40] has pointed out earlier that low leach rates of
complex nuclear waste glasses are due to Type IIIB surfaces, which
are composed of multiple layers of oxides, hydroxides and hydrated
silicates resulting from a sequence of solution-precipitation
reactions between the glass surface and leaching solutions (Fig.
2-3). A number of alkali borosilicate nuclear waste glasses that
exhibit Type IIIB surface behavior have elemental leach rates as low


21
CM
O
TYPE I
Original glass solution interface
\ BULK -
z:
o
o
CO
W
o
Inert glass
z
o
)-
Z)
to
TYPE II
-Selective loachlng
-BUU<-
Protsctive film
on glass
-OISTANCE-
-Dl STANCE-
TYPE ni e
Fig. 2-3. The five types of glass surfaces and six surface
conditions resulting from glass-environment interactions
(adapted from [40]).


22
2
as 0.02 to 0.2 g/m .day with a time dependence of static leaching of
to t0'^ or less after 28 days at 90C.
In this work terms such as selective leaching and congruent
dissolution are used in discussing the glass leaching mechanisms.
Selective leaching includes ion exchange of the mobile species in the
glass and selective dissolution of glass matrix, structural or
network species with or without precipitation. Ion exchange involves
a process in which exchange between mobile species such as Na from
the glass and hydrogen or hydronium ions from the solution occurs.
Ion exchange can also occur between Ca, Mg and K in ground water with
mobile species from the glass. During this process, the remaining
constituents of the glass are not altered. As mentioned in Chapter
I, the structural network of the borosilicate glass is provided by
[SiOij]14 [B0|]^ and [BO^]^ polyhedra. Since different glass
network formers dissolve at different rates, selective dissolution of
matrix, structural or network species is usually observed with a
multicomponent glass containing two or more network formers. This
may or may not be followed by precipitation depending on the
composition of glass and solution. Congruent dissolution occurs when
the species comprising the glass are dissolving into solution in the
same ratios as they occur in the bulk glass. Without precipitation
the composition in the glass surface is not changed by congruent
dissolution. However, large dimensional changes often accompany such
kinds of corrosion. Congruent dissolution may be followed by
precipitation after certain less soluble species approach
saturation. In this case, the composition of the glass surface


23
changes away from that of the bulk glass and less soluble
constituents are enriched at the altered layer.
Although a short (several days) period of predominant alkali-
hydrogen ion exchange may occur for Type IIIB glasses, the dominant,
long-term mechanism controlling corrosion is a combination of more or
less selective dissolution of glass matrix followed by
precipitation. The extent of matrix dissolution and onset of surface
and inner precipitation will depend on the time required for various
species in the glass to reach saturation in solution. Saturation of
a certain species will be a function of the initial solution pH,
concentration of alkali species in the glass and their rates of
release which change the solution pH, temperature, initial
concentration of that species in the solution, the ratio of glass
surface area to volume of leachant (SA/V) which influences solution
concentration, and flow rate which also affects solution
concentration.
The theoretical basis for Type IIIB glasses is the investigation
of Grambow which predicts the formation of a series of insoluble
reaction products on glass surfaces [41]. He concluded that reaction
of the matrix is the fundamental process that occurs in the leaching
of alkali borosilicate nuclear waste glasses. He pointed out that
without solubility restrictions, congruent dissolution occurs at all
pH values and leachant compositions. That is, the glass dissolves
congruently at a rate proportional to kt1. Even after saturation has
occurred with respect to a certain species, the glass can continue to
dissolve congruently with simultaneous precipitation of that species.


24
When solution saturation of species "i" is reached, there is no
longer any driving force for that species to leave the glass
surface. Consequently species i will accumulate at the glass-
solution interface as the matrix dissolves. If matrix dissolution
releases alkali ions, as will be the case for most glasses, there
will be a concomitant rise in pH proportional to the flow rate or
SA/V of the system. An increase in pH can have several simultaneous
effects on the glass, the solution and the glass-solution
interface. At the new pH, a second species "j" may reach solution
saturation and subsequently be retained in the glass surface along
with species "i." The extent of apparent incongruent dissolution of
the glass is thereby increased. The sequence of events that occurs
is predictable, based on the solubility limits of each species at a
given pH, as shown by Grambow [41].
Figure 2-4 summarizes the behavior of various elements
considered by Grambow [41]. Here, the ratio NS^/NLg^ is used to
present the solubility limits of these elements in solutions of
different pH. The normalized solubility NSi (g/m ) is given by
NS.
i
C
f.
i
i, sat
SA/V
(2-2)
where ^ = the solubility limited concentration in the leachate
at the specified conditions, g/L;
f^ = the mass fraction of element i in the glass;
2
SA = the specimen surface area, m ;
V = the volume of the leachage, L;


25
p
NLg^ = the normalized elemental mass loss of Si, g/m as
defined in equation (2-1).
Therefore, in static leach tests, nuclear waste glass containing Fe
oxides should concentrate Fe within surface layers. Zinc, Nd, Sr and
Ca should be concentrated as well in nearly neutral or slightly
alkaline solutions with Na and B depleted.
The low overall leachability of many nuclear waste glasses over
a pH range from 4.5 to 9.5 is a consequence of the formation of the
multiple barrier (Type IIIB) films. Figure 2-5 is a plot of Si
leachability for an SRL composite waste glass immersed in a 5~day
static 23C solution buffered to various pH values from 3-5 to 10.7
[42]. These data show that over the pH range expected for repository
ground waters, indicated by arrows, glass leachability is low, since
the formation of less soluble reaction products lowers the solubility
of silica in the solution [43].
One thing that Grambow did not explain with his leach data is
why NL^/NLg^, where i is Ca, Fe, Zn, Nd or Ce, is usually larger than
1 when the solubility restrictions are removed. If NL^/iNLg^+NLg)
had been used, it may be much closer to 1, since BgO^ is also a glass
network former.
Molecular Mechanism of Aqueous Dissolution
In discussing the molecular mechanism of aqueous dissolution of
alkali borosilicate glasses, Grambow [44] extended the idea of
Aagaard and Helgeson [45] on the pure silica-water interactions, and
interpreted the observed saturation effects as a local surface
equilibrium process involving the critical activated surface


26
Fig. 2-4. Ratio of normalized solubility to NLg (20 g/m2) for
CaC03> SrCOg, Nd(0HK, Fe(0H)3 and Zn(0H)2 in MCC-1 28-day
test at 90C in solutions of different pH (adapted from
[41]).


Leachabity, g/m
27
6
pH
Fig. 2-5. The Si leachabity of a borosilicate glass immersed in a
5-day static 23C solution buffered to various pH values
(adapted from [M2]).


28
complex: for every complex desorbed from the glass matrix, another
complex is adsorbed from solution (equal forward and back
reactions). However, compared to the silica-water system, the waste
glass-ground water system is much more complex; dozens of elements
are involved in the system in addition to a dependence on variables
such as Eh, pH, ground water composition. Furthermore, when using
the approach of Aagaard and Helgeson, it is assumed that rates of
hydrolysis are controlled primarily by reaction kinetics at activated
sites on the surface of glass and not by diffusional transfer of
material through a leached outer zone or a coherent surface layer of
reaction products. Grambow [44] assumes that there exists a critical
surface complex whose desorption controls the mobilization of various
glass constituents. In contrast to saturation of Ca, Fe, Nd, and
others, saturation of silica in solution has a major effect on the
corrosion rate. Silica is the dominant constituent of the activated
complex and, according to Grambow's arguments, its desorption (as
silicic acid) from the glass network will limit the rate of release
of other glass constituents, even when these elements are not
solubility limited in solution. At saturation, condensation of
silanol groups will stabilize the glass network against further
attack of aqueous species.
Recent data from flow tests [46] as well as other investigations
[47] indicate the importance of considering diffusion in the leached
layer in addition to the reaction kinetics at the activated sites on
the surface of glass. Data from flow tests [46] indicate an initial
increase to a maximum of the solution concentration of various glass


29
constituents followed by a decrease with time. Grambow et al. [47J
speculate that the leaching is controlled by the transport of silicic
acid through a growing surface layer, as shown in Fig. 2-6. In this
figure the saturation concentration is not a constant because the pH
in the surface layer varies with time. Surface layer diffusion
results from the difference between the silicic acid concentration in
the bulk solution and at the surface layer. In the surface layer,
saturation is reached after 200 days, whereas the solution
concentration is still far below saturation. The same general trends
are observed for other elements such as Na and B [46].
Systems Interaction Tests
A comprehensive systems interaction study was performed by Wicks
et al. [48] in which they compared leaching behavior of a defense
waste glass in deionized water, ground water and both waters
containing rocks. Their analyses were based on the concentrations of
species in solution and did not take into account the species
adsorbed onto solids present in the leaching vessel. They found that
the presence of salt (from Carlsbad and Avery Island), basalt, shale,
granite and tuff all slightly decreased the concentrations of glass
species in solution compared to those obtained when deionized water
was used alone. Similar results were found with synthetic ground
waters and with the same water containing the various rocks. Actual
ground water yielded results comparable to the MCC reference waters
and synthetic ground waters.
Clark and Maurer [49] have investigated the effects of several
types of rocks, including basalt and granite, on the leaching of a


Si CONCENTRATION (mg/I)
30
_r0ZZr.ZZZ=nmnzr_ 1
0 100 200 300 400
TIME (d)
Fig. 2-6. Calculated silicon concentrations in the surface layer and
bulk solution based on the surface layer diffusion and pH
as a function of leaching time. A diffusion coefficient
of 10 9 cm2/sec was assumed [46,47].


31
borosilicate glass. With the possible exception of granite, the
combination of glass and rocks in the same leaching vessel did not
appear to have any significant effects in a 28-day test.
Brine solution generally decreases the rate of glass corrosion
[48,50,51], with the possible exception of Sr, Ce and similar
elements [35]. In the brine solutions, a protective magnesium
chloride complex forms on the glass surface. Exposure of nuclear
waste glass to tuffs results in a small decrease in corrosion rate,
perhaps due to a buffering effect [48]. Autoclave tests of basalt-
glass interactions [52] and granodiorite-glass [53] interactions show
a decreasing rate of attack.
McVay and Buckwalter [54] investigated the effect of iron on
nuclear waste glass leaching. They found that the presence of
ductile iron in deionized, tuff and basalt ground waters containing
PNL 76-68 borosilicate glass caused significant changes in the
leaching characteristics of the glass. Formation of iron silicate
precipitates effectively removes many elements from solution and
therefore inhibits the saturation effects which normally cause
decreases in elemental removal rate. Thus, basalt and tuff ground
waters behave similarly to deionized water in the presence of ductile
iron. The precipitates also retard saturation effects, resulting in
high sustained leach rates and thus greater total elemental removal
from the glass. A synergistic effect occurs between the two
materials. The iron enhances glass leaching and the glass enhances
iron corrosion.


32
The presence of a radiation field during storage and its effect
on glass leaching is another consideration. Radiation may affect the
glass-water system in several ways. Gamma radiation has been found
to result in approximately three- to seven-fold increases in the
leach rates of borosilicate glasses [55,56]. As reported by McVay
and Pederson [55], some of the enhancement is due to nitric acid
formation from air radiolysis in the presence of water. Nitric acid
appears to preferentially attack zinc and lanthanides, both of which
normally build up on the surface of the PNL 76-68 glass when leached
in nonacidic solutions. The change of the solution chemistry by
gamma radiation and generation of reactive species such as OH- from
water radiolysis also appear to be important. The principal effect
of water radiolysis products is the increased silicate dissolution.
The leaching behavior of the radioactive glass has been
investigated in comparison to that of the simulated glass [57]. In
this case it was found that radiation, due to the low dose rate with
the radionuclides (0.594 Ci per specimen), does not affect
significantly the leaching rate. This conclusion includes the
effects of radiation damage to the glass itself and the interaction
of the radiation field from the glass with the water and air.
Profiles of Pu and U behave similarly during leaching, both being
enriched in the surface of the glass. Leaching of radioactive glass
results in loss of B, Na, Li and Mo with about the same depth of
leaching. The leaching mechanisms appear to be similar for
radioactive and nonradioactive glasses [57].


33
Burial Studies
Burial studies were started in the late 1950s and early 1960s.
Merritt and Parsons [24,25] pioneered two tests of high-level waste
(containing real radionuclides) incorporated into nepheline syenite
glass and buried in contact with ground water for 15 years at Chalk
River, Canada, at ambient temperature. Fletcher [26] conducted
burial experiments of waste glass samples in England for up to 9
years. Although the field tests were not performed under actual
repository conditions, they did provide an approximation to a
potential repository. Preliminary results from burial experiments
[24,25] have shown that glasses leached at much lower rates under
repository-like conditions than under laboratory conditions. As an
example, the observed field leach rate from the Canadian burials was
over 200 times lower than the lowest leach rate reported in the
laboratory [24,25]. The authors attributed about 1/5 of this
difference to the lower aggressiveness of ground water over distilled
water used in the laboratory experiments and to its lower temperature
(6C in the field compared to 25C in the laboratory). The remainder
of the difference was attributed to the formation of a protective
surface layer.
The leaching performance of a waste glass depends on the
environment under which it is tested. In a repository, the system
variables ultimately controlling the environment to which the waste
glass is exposed include geology, engineered waste package
components, initial ground water chemistry, temperature, pressure,


34
radiation field, water contact time and flow rate through the
repository.
The most extensive and systematic field tests began in 1982 and
involve deep burial (350 meters below surface) in granite in the
Stripa mine, Sweden [16-20]. The boron depletion depths of glass
ABS 39* and 41* ranged from 0.2 pm to 15 pm, depending on composition
and the type of material to which the glass was exposed after 1 year
of burial at 90C. At the glass/glass interface, both glasses showed
a depletion of Na, Cs and B, but for the more corrosion-resistant
glass, the lower depletion depth was ascribed to the formation of a
thin (0.2 pm) coherent and dense outer layer, enriched in Mg, Ca, Sr,
Ba, Zn, Al, Fe and Si, which impedes both the ion exchange and
network attack of the bulk glass underneath. The presence of
bentonite increased the boron depletion depth up to 1 year by a
factor of approximately 5, whereas granite decreased this depth by
about 2 times. This behavior is attributed to bentonite serving as a
semi-infinite ion exchange medium where Ca from the bentonite is
replacing Na, Li and B from the glass [19]. In contrast, the small
congruent solubility of granite seems to augment the glass in
reaching solubility-limited leaching [21].
Another in situ test was initiated in 1986 and involves burial
in a clay formation in Mol, Belgium [22]. A number of simulated
waste forms (including HLW glasses and glass-ceramics) have been, or
will be, buried at the site. Their corrosion rate will be measured
* Swedish alkali borosilicate (ABS) and nuclear waste glasses.


35
in two environments susceptible to contact the radioactive waste
during its geological storage in a clay formation: host clay and a
humid atmosphere loaded with clay extracts. The tests, with total
exposure times of 6 years, will be carried out at various
temperatures, 15, 50, 90 and 170C.


CHAPTER III
RESEARCH OBJECTIVES, APPROACH AND SUMMARY OF CONCLUSIONS
Research Objectives and Approach
The primary objectives of this study were (1) to evaluate the
leaching behavior of selected nuclear waste glasses in a realistic
repository environment, (2) to develop a characterization methodology
for comparing field data with laboratory data and (3) to assess
leaching mechanisms and to correlate the mechanisms observed in
laboratory-leached vs field-leached specimens.
In order to achieve these objectives, both field experiments and
laboratory simulation tests were conducted. The field tests involved
long-term (up to 31 months) deep burial (350 m below surface) in
granite in the Stripa mine in Sweden. Two configurations of samples
were used. One was a 32 mm in diam. x 35-mm long minican where an
alkali borosilicate glass with simulated HLW was cast into stainless
steel. The second configuration was the so-called "pineapple
slices," 51-mm in diam. x 5-mm thick, which resulted in a variety of
glass/repository materials interfaces. Two temperatures, 90C and
the ambient temperature (8-10C), were used to simulate the
repository conditions during and after the thermal period of storage,
respectively. Comparisons were made of six alkali borosilicate
simulated nuclear waste glasses, including three American Savannah
River Laboratory (SRL) glasses and three Swedish alkali borosilicate
36


37
(A3S) glasses. Different glass/repository materials interfaces were
provided to investigate effects of these materials on glass leaching.
In the laboratory, methods were designed to consist of both
static and single-pass low-flow tests, using granite rock cups as
leach vessels and Stripa ground water in an attempt to closely
simulate the repository-like conditions in Stripa.
Several research tools, including solid surface analysis and
solution analysis techniques, were used in combination. These
provided a direct evaluation of nuclear waste glass leaching under
various test conditions.
Major Conclusions
1. A significant compositional effect on glass leaching was
observed under burial conditions. The leach rate expressed by the
annual boron depletion depth was inversely correlated with (SO2 +
A^O^/O^O +02O2) wt ratio in the glasses; F^O represents the alkali
oxides.
2. Accelerated attack during the first year in the presence of
bentonite appears to be a transient effect. The presence of
stainless steel, Cu and Ti does not have much effect on glass
leaching.
3. The leach rates of buried samples based on boron depletion
at 90C ranged from 0.3~3 pm/year for the glass/glass interfaces
investigated. These values are at least two orders of magnitude
lower than those for glasses leached using MCC-1 static leaching
procedures and deionized water.


38
4. Comparison of the laboratory simulation results with field
test results indicates that glass leaching mechanisms were similar
under both test conditions.
5. A model, based on glass structure and thermodynamic
considerations, was proposed to better describe alkali borosilicate
glass leaching than the recent model proposed by Grambow.
6. The results show that Stripa burials combined with
laboratory simulations are unique experimental designs which have
provided useful information regarding nuclear waste glass leaching.
This work has served as a model on which design and development of
the most recent burial test programs are based.


CHAPTER IV
MATERIALS AND METHODS
Glass Compositions and Characterization
Burial Samples
Six alkali borosilicate simulated nuclear waste glass
compositions were used in the burial experiments. They included
three American SRL glasses and three Swedish ABS glasses. Their
compositions are listed in Table 4-1 .
Frit 131 and frit 165, which were designed to contain the
Savannah River Plant (SRP) nuclear waste, were used to prepare SRL
glass samples. Two glasses containing 29.8 wt$ and 35 wt TDS*
waste, respectively, were prepared from frit 131. Another SRL glass
was prepared from frit 165 and contains 29.8 wt? TDS waste. ABS 39
and 41, developed and produced by Dr. T. Lakatos, Swedish Glass
Research Institute, Vaxjo, Sweden, contain 9% by weight simulated
fission product oxides. These glasses are similar to the COGEMA
glass selected for vitrification of commercial HLW in LaHague,
France, operations [23]. ABS 118 contains 11.25 wt simulated
fission product oxides and has a composition very close to that of
the future COGEMA glass.
The glass frits were premelted from chemicals using standard
procedures. The simulated SRP waste was mixed with the frit before
* See the notes in Tables 4-1.
39


40
Table 4-1.
Nominal Waste Glass
Burial.
Compositions
(wt?)
Used in the
Stripa
SRL 131 +
SRL 165 +
SRL 131 +
Component
29.8? TDS+
29.8? TDS+
35? TDS+
ABS 39
ABS 41
ABS 118
From glass
frit
Na20
12.4
9.1
11 .5
12.9
9.4
9.9
L i ^0
4.0
4.9
3-7

3.0
2.0
ZnO




3.0
2.5
MgO
1.4
0.7
1.3



AlpOq



3.1
2.5
4.9
B20^
10.3
7.0
9.6
19.1
15.9
14.0
Fe^^



5.7
3.6
2.9
L
0.4

0.3



Si02
40.6
47.7
37.6
48.5
52.0
45.5
Ti02
0.7

0.7



Zr02
0.4
0.7
0.3


1 .0
uo2



1 .7
1.6
0.9
P25





0.3
Cr22





0.5
NiO





0.4
CaO

--

--
4.0
From simulated waste
Fe^^
13.4
13.4
15.8



Mn02
3.9
3.9
4.5
0.78
0.78
0.97
Zeolite** 2.9
2.9
3-4



A^O^
2.7
2.7
3.2



NiO
1 .6
1 .6
1.9
0.37
0.37
0.47
Si02
1 .2
1 .2
1 .4



CaO
1 .0
1 .0
1.2



Na20
0.9
0.9
1 .0



Coal
0.7
0.7
0.8





Table 4-1.--continued.
Component
SRL 131 +
29.8% TDS+
SRL 165 +
29.8% TDS+
SRL 131 +
35? TDS+
ABS 39
ABS 41
ABS 118
Nc^SOjj
0.2
0.2
0.2



Cs2C03
0.1
0.1
0.2



SrCO^
0.1
0.1
0.2



U38
1 .1
1 .1
1.3



Cs20



0.89
0.89
1.11
SrO



0.26
0.26
0.33
BaO



0.46
0.46
0.58
Y23



0.15
0.15
0.19
Zr02



1.29
1.29
1.62
Mo03



1 .65
1 .65
2.06
Ag20



0.01
0.01
0.01
SnO



0.02
0.02
0.02
Sb ^0^



0.004
0.004
0.005
L



0.72
0.72
0.90
Nd203



1.22
1 .22
1 .53
Pr203



0.38
0.38
0.48



0.76
0.76
0.95
CdO
--
0.03
0.03
0.03
Total
100.0
99.9
100.0
100.0
100.0
100.1
+ TDS waste as received from SRL contained Fe203, Mn02, zeolite,
A120o, NiO, Si02, CaO, Na20, Coal and Na2S0^. This waste was also
doped with Cs, Sr and U.
** Zeolite contains (in wt?) 67.2 Si02, 19.3 A1203> 6.3 Na20, 3.4
Fe203> 2.8 CaO and 1.0 MgO.


42
vitrification. The mixture was fused at 1150-1200C for 2-6 hours
and annealed at 500-525C for 1 hour.
Two sample configurations were used: (1) minicans and (2)
pineapple slices. The minicans were made by casting the molten glass
in stainless steel rings 3 mm in diameter by 35 mm long. After
annealing, a hole 200 mm in diameter was drilled through the center
of each minican. Both surfaces of the minicans were polished to a
6-pm finish with diamond paste. Pineapple slices were prepared by
casting cylinders 51 mm in diameter by 80 mm long in molds containing
center carbon posts. Sections 5 mm thick were sliced from the
annealed cylinders and the center posts were removed. One side of
each pineapple slice was polished to a 600-grit (-17 pm) surface
finish while another side was kept in as-cut condition for easy
identification of the glass interfaces after burial. Figure 4-1
shows the pineapple slices of glass, granite,* stainless steel, Ti,
Pb and compacted bentonite** before burial.
Before burial, each sample was subjected to two types of surface
analyses: (1) optical microscopy and (2) Fourier transform infrared
reflection spectroscopy (FT-IRRS). Four to six spots on the polished
surface of each pineapple slice and two spots each on both sides of
the minican were analyzed using these two techniques of surface
* The granite was obtained from Stripa, Sweden.
** The bentonite was obtained from Wyoming. The compacted bentonite
was made by means of isostatic compaction under 100 MPa of
pressure. This is a so-called sodium bentonite whose main
constituent (90 wt) is montmorillonite.


-fcr
UJ
Fig. 4-1. Pineapple slices of glass, granite, stainless steel, Ti, Pb and compacted bentonite before
burial.


44
analyses. In addition, each sample was weighed before burial. Table
4-2 is the sample matrix of the burial experiments.
Glass Quality
Fourier transform infrared reflection spectroscopy (FT-IRRS) was
used as a nondestructive analytical tool for characterization of
glass surfaces prior to the burial. One objective of this
statistical analysis was to determine the relationship between the
FT-IRRS spectra and glass composition used in the Stripa burial. A
second objective was to check if there were any appreciable
variations in composition and/or surface finish conditions among
samples of the same glass formulation. SRL glass samples were used
in this study. The FT-IRRS spectra were obtained on 4 to 6 spots
along the diameter of each glass sample.
Table 4-3 lists the statistical variations of the FT-IRRS
analysis for the SRL glasses. Figure 4-2 shows the representative
spectra of SRL glass pineapple slices before burial. It is observed
that both the wavenumber and the intensity (integrated area under the
curve) for the broad peak containing Si-0 stretching vibrations
increase with increasing SO2 content in the glass composition, i.e.,
in the order of SRL 131 + 35% TDS, SRL 131 + 29.8? TDS, SRL 165 +
29.8% TDS. Range of variation in peak position for the same
composition was 0.4-0.8?. The standard deviation of peak position
for the SRL 131 + 29.8% TDS glass slices was the largest (0.8?) due
to glass heterogeneities contained in a few samples of this
composition. On the other hand, the peak intensity and integrated
area under the spectra vary more than peak position because peak


45
Table 4-2. Sample Matrix of the Stripa Burial Tests.
Time
(month)
SRL 131 +
29.8? TDS
SRL 165 +
29.8? TDS
SRL 131 +
35? TDS
ABS 39
ABS 41
ABS 113
Minicans*
It
1
90C
90C



--
3
90C
90C




12
90C
90C



--
24
8-10,90C
8-10,90C




Pineapple Slices**
1
90C
90 C
90C
90C
90C

2





90C
3
90C
90C
90C
90C
90C

4





90C
6
90C





7





90C
12
8-10,90C
90C
90C
90C
90C
90C
24
8-10,90C
8-10,90C
8-10,90(


31
--
90C
90C
--
* Including glass/glass and glass/bentonite interfaces.
** In the case of SRL glasses, glass/glass, glass/bentonite,
glass/granite, glass/Ti and glass/stainless steel were included
with extra two interfaces, glass/Cu and glass/Pb for 1-month
burial; all ABS glasses included glass/glass, glass/bentonite,
glass/granite, glass/Ti, glass/Cu and glass/Pb interfaces.


46
Table 4-3. Variations in Spectral Characteristics of SRL Waste
Glasses.
Glass
Peak
Location
(cm 1 )
Peak*
Intensity (?)
Integrated Area
(Relative Value)
Minicans
SRI 131 +
29.8% TDS
989

ij**
21.26 1.67
5.22
SRL 165 +
29.8% TDS
996

6
23.12 1.53
5.58
Pineapple
Slices
SRL 131 +
29.8? TDS
980

8
21 .71 2.42
5.16
SRL 165 +
29.8? TDS
990

6
22.82 2.34
5.44
SRL 131 +
35? TDS
974

6
20.16 2.63
4.67
* The compound peak containing Si-O-Si stretching vibrations at
800-1200 cm-1 was used in the statistical analysis.
** Mean and standard deviation.


REFLECTANCE (%)
30.0
WAVENUMBERS
Fig. 4-2. Representative FT-IRRS spectra of SRL glass pineapple slices before burial.


intensity is sensitive to variations in the surface roughness due to
polishing. All these results show that the glass samples except
those containing crystallites are homogeneous compositionally but the
surface polishing conditions have relatively wide variations.
Laboratory Samples
In laboratory simulation tests, most of the glass samples were
made from two similar glass formulations, SRL 165 + 29.8? TDS and
Black Frit 165-Mobay (see Table 4-4). The same melting procedures as
for the burial samples were followed in making the laboratory
glass. The glass melt was cast into a graphite mold. The glass bars
were annealed at 500C for 1 hour, then furnace cooled.
After cutting from glass bars, samples were polished on all
surfaces up to 600 grit with SiC papers. After cleaning, each sample
was subjected to two kinds of surface analyses: optical microscopy
and FT-IRRS. All samples were weighed before corrosion.
Stripa Field Tests
Sample Assemblies, Minicans and Pineapple Slices
The minicans and the pineapple slices, granite slices, compacted
bentonite slices, stainless steel, Ti, Pb and Cu coupons were
0
assembled at the University of Lulea, Sweden, to provide a wide range
of glass/repository materials interfaces. Minicans were designed to
closely simulate a waste package in a disposal hole. Minicans and
compacted bentonite coupons were stacked together to provide
glass/glass and glass/bentonite interfaces (Fig. 4-3). Sleeves of Pb
and Ti or Cu overpacks were placed around the steel wall of the
minican and a bentonite sleeve separated the waste package from the


49
Table 4-4. Nominal Composition
of Black Frit 165
Mobay Glass.
Component
Wt %
Si02
55.61
Fe22
11.34
Na20
10.44
B2O3
7.23
L i ^0
4.82
A1203
4.22
Mn02
2.11
CaO
1.10
NiO
0.90
MgO
0.70
Zr02
0.90
F
0.14
Cl
nil
Pb
nil
k2o
0.14
Ti02
0.23
BaO
0.07
ZnO
0.05
Total
100.00
Note: Glass was supplied by
Savannah River Laboratory, Aiken,
SC. This is a similar formula
tion to SRL 165 + 29.8$ TDS.
However, it contains more Si02
and less Fe203-


50
Fig. 4-3. A minican assembly


51
walls of the borehole. A typical pineapple slice assembly before
burial is shown in Fig. 4-4.
The SRL glasses included seven pineapple slice assemblies with
different sample stacking sequences (Fig. 4-5 and Table 4-2) and five
minican assemblies with the same sample stacking sequence (Table
4-2).
All the Swedish ABS glass assemblies had the same stacking
sequence to provide six different interfaces (see the footnotes in
Table 4-2). Thus, 35 glass/repository materials interfaces were
involved in these Stripa burial tests with six alkali borosilicate
simulated nuclear waste glass compositions.
Stripa Repository
The Stripa abandoned iron mine was chosen as an underground
field laboratory where the major rock formation is a massive, grey to
light red, medium-grained granite. The mine is located in central
Sweden. The massive and compact nature of granite makes it very
impermeable to water. The hard rock formation has great structural
strength and resistance to erosion or other disruptive events.
Hence, nuclear waste glass placed deep in granite is very unlikely to
be disturbed by climatic or geological events, or by accidental human
intrusion [58].
Table 4-5 lists the average major/minor chemical and mineral
constituents of the Stripa granite. There are several fracture
systems. The majority of the fractures are closed and filled mainly
with chlorite but occasionally with calcite. This mine provides an
environment which closely simulates an actual granite repository and


52
Fig. 4-4. A typical pineapple slice assembly.


Fig. 4-
Pre- Burial
6 iz
10* C
ai
LO
5. Seven preburial pineapple slice assemblies with different sample stacking sequences for SRL
simulated nuclear waste glasses.


54
Table 4-5. Average Major/Minor Chemical
and Mineral Constituents in
Stripa Granite.
Oxide
Wt %
Si02
74.7
Al^jO^
13.2
Fe20n
1 .6
FeO
NR a
MgO
0.20
CaO
0.6
Na20
4.0
k2o
4.6
Ti02
0.05
P25
NR
MnO
0.03
BaO
0.02
h2o
NR
OJ
o
o
NR
Mineral
Grey
Red (vol %)
Quartz
33
44
K or Na Feldspar
24
12
Plagioclase
35
39
Biotite
<1
NR
Muscovite
<1
2
Chlorite
<1
3
Adapted from [59].
a NR = not reported.


55
is thus ideal for conducting glass corrosion experiments. The
location within the Stripa mine where the samples were buried is
shown in Fig. 4-6. This is about 345 meters below the surface. The
holes into which the samples were placed were about 3-m deep and
56 mm in diameter, which were filled with ground water from the mine
before the samples were placed to a depth of about 2.5 m (see Fig.
4-7). The arrow indicates holes for 1-month, 90C specimens. The
ground water composition and pH prior to burial as measured in the
recent study is given in Table 4-6. Table 4-7 lists the ground water
composition and pH found in literature [60],
Burial and Retrieval
All the sample assemblies were buried in the boreholes at the
345-m level below the surface. Heater rods were placed in the center
20-mm holes on the samples designed to be maintained at 90C (Fig.
4-6) to simulate the thermal period of -300 years. Without the
heating elements, glass samples were tested at ambient mine
temperature (8-10C), which is expected to be the temperature of a
canister in a real repository after about 300 years. As shown in
Fig. 4-7, a rubber seal was used to prevent water intrusion from the
floor of the mine. Water entering the hole had to permeate through
the granite.
Assemblies were retrieved at specific intervals over a 3_year
period (see Table 4-2). After removal of the burial assemblies from
the boreholes, they were wrapped in plastic until disassembled and
analyzed. Water from the boreholes was analyzed prior to and after
the assemblies were removed.


56
BURIAL SITE AT STRIPA
GRANITE
SHAFT
MINE TUNNEL
CARP El
HOLE
Fig. 4-6. Location within Stripa where SRL samples were buried.
This is about 345 m below the surface. The holes into
which the samples were placed are about 3~m deep and 56 mm
in diameter. They were filled with water from the mine
before the samples were emplaced. The arrow indicates
holes into which the 1-month, 90C specimens were
placed. Samples were placed in the hole to a depth of
about 2.5 m.


57
SCHEMATIC OF BURIAL ASSEMBLY AT STRIPA
GRANIU
345 meter
Section of Sdmplr
(Pff Burial)
Fig. 4-7. Diagram illustrating the position of the samples in the
Stripa mine during burial. A pineapple slice and minican
are also shown, along with a photograph of the 1-month,
90C assembly immediately after removal from the borehole.


Table 4-6.
Ground Water Composition and pH Measured in This Study within the 1-month Test Hole at
Stripa. Concentration mg/L.
pH Li B Na Al Si Mn Fe Zn Sr Mo La Nd ^
CO
Before Burial 8.1 0.16 <1 30 0.02 7.7 <0.1 <0.1
0.30 0.16 <0.01


59
Table 4-7.
Ground Water Composition for
the Stripa Granite, Literature
Values.
Anions
mg/1
hco3"
Cl"
so42~
15.4-78.7
52-283
2.7-1.9
f"
NRa
Cations
Ba2+
Ca2+
Fe3 +
NR
10-59
0.02-0.24
Li +
Mg2 +
K+
NR
0.5
0.2-5.4
Si02
Na+
Sr2+
11 .0-12.8
43-125
NR
pH
8.85-9.75
Total Dissolved 200-230 (330-410 m)b
375-510 (below 700 m)b
Adapted from [60].
a NR = not reported.
b Depth of sample below surface.


60
The measured flow rates through the boreholes near those where
the assemblies were located were approximately 1 L/year (0.1 mL/hr)
[61]. The glass surface area to ground water volume ratios (SA/V)
were estimated to be >1 cm 1 and were most likely different from spot
to spot on some of the samples due to different water accessibility
at the glass interfaces. The calculated SA/V ratios were low, about
0.6 cm 1 for the pineapple slices and 0.06 cm 1 for the minicans.
This was based on the volume of water below the rubber seal and the
total surface area of the glass in the hole.
The postburial procedures consisted of careful disassembling,
soaking in deionized water for no more than 5 min to remove excess
bentonite, if present, and two to three 5-min ultrasonic cleanings in
acetone or absolute ethanol. The samples were air-dried and placed
in a desiccator until analyzed.
Disadvantage of the Burial Test Method
The primary disadvantage of the Stripa burial is that it was not
possible to calculate leach rates based on the data of solution
analyses. This is because, when the samples were taken out of the
borehole, the ground water above the rubber seal ran into the lower
part of the borehole where the sample assembly was positioned. In
addition, other contaminants may be present in the ground water. All
these make the leach rate calculation based on the solution analysis
data meaningless. Therefore, surface analyses had to provide the
primary evaluation method for assessing glass performance and for
comparing field- and laboratory-corroded specimens. As will be
mentioned, the Materials Interface Interactions Tests


61
surface/solution analysis (MIIT-SS) effort contains an improvement
over the Stripa burial in that solution analyses will be obtained
[62].
Similar Tests Being Used in MIIT Studies at WIPP
The Materials Interface Interactions Tests (MIIT) is a series of
experiments that will assess the performance of simulated SRL waste
glass along with a variety of additional simulated waste glass
compositions in the presence of various proposed canister, overpack
and backfill components, in the salt geology at the Waste Isolation
Pilot Plant (WIPP) [62]. Design and development of the MIIT tests
were derived from the experience obtained through in-situ testing of
over 100 simulated SRL waste glass samples buried in Stripa granite,
Sweden. The MIIT in-situ testing program represents a "second
generation" of the Stripa tests. The MIIT studies consist of two
parts, MIIT-MI (multiple interactions), and MIIT-SS (surface/solution
analysis). The MIIT-MI effort is similar to the Stripa experiments
and involves glass performance as a function of a variety of proposed
package components, predominantly by surface analyses. The MIIT-SS
effort represents a significant improvement over the Stripa burial
experiments in that solution analysis will also be obtained for
simplified interactions, and time-dependent data will be obtained
from single boreholes. Only pineapple slice assembles will be
utilized. All tests will be conducted at 90C. Samples will be
removed from the mine at time intervals of 6 months, 1 year, 2 years
and 5 years.


62
Laboratory Tests of Simulated Corrosion
In order to simulate the actual repository conditions, two sets
of laboratory leaching tests were conducted. In one set, a modified
MCC-1 static leach test method was used for SRL 165 + 29.8? TDS glass
with two different glass surface area-to-volume of leachant (SA/V)
ratios, 0.1 and 1.0 cm 1. The leachant was selected from one of the
following: deionized water, Stripa ground water and Stripa ground
water saturated with glass powders of the same composition as the
bulk specimen at 90C for 14 days. Prior to immersing in the
leachant, each specimen was ultrasonically cleaned in either reagent
grade acetone or absolute ethanol for 3 times, 5 min each. The
samples were suspended inside either a PFA Teflon* (60 ml capacity)
corrosion cell and then placed inside a constant Blue M** convection
oven as shown in Fig. 4-8.
In another set of so-called "rock cup tests," a Stripa granite
cup was placed in each PFA Teflon container to simulate granite
repository conditions. The granite was obtained from boreholes in
Stripa, Sweden, and the granite cups were made by Diversified Machine
Works, Post Falls, Idaho. A diamond drill was used to drill a hole,
3.2 cm in diam. by 3-8 cm deep in each granite cylinder, 4.4 cm in
outside diam. by 4.9 cm high. Monolithic glass samples of Black Frit
165-Mobay (see Table 4-4) were placed in the cup. A certain volume
of ground water was filled both inside and outside the cup. Some of
* 0102-53 MOD PFA Teflon jar, Savillex Corp., Minnetonka, MN.
** Model OV-490A-2, Blue M Co., Blue Island, FL.


ENVIRONMENTAL TESTING -SYSTEM
Fig. 4-8. Schematic of experimental configuration of static leach test.


64
the cups contained stainless steel (316 L) wires used for supporting
the glass specimens. The rock cups were soaked in ground water for 2
days, then air dried.
Both static and flow test conditions were used in the rock cup
tests. The SA/V ratio was 1.0 cm 1 in the cells. In the case of
static leaching, the surface area of glass sample was about 14 cm .
A stainless steel (316 L) wire, 0.1 cm in diameter by 18 cm in
extended length was contained in the rock cup. All glass samples and
stainless steel wires were cleaned ultrasonically 3 times, for 5 min
before leaching with absolute ethanol. A corrosion cell for the rock
cup static test is similar to that for the flow test shown in Fig.
4-9 but without the fittings in the lid of the Teflon container.
In the rock cup flow test, low flow rates, 0.1-0.3 mL/h, were
used. This single pass continuous flow test method was similar to
the MCC-4S procedures [27]. A stainless steel (316 L) wire, 0.1 cm
in diameter x 36 cm in extended length, was contained in the rock
cup. The glass surface area was about 26 cm The procedures of
sample preparation and granite cup cleaning were the same as in the
static test. A flow leaching vessel and the experimental set-up are
shown in Figs. 4-9 and 4-10. Only the leachant within the cup was
forced to flow using a Peristaltic cassette pump.* The flow rate of
the ground water was controlled to 10? of the set value. The ground
water was not preheated in the reservoir. Since the flow rate was
low and the ground water prior to being introduced into the leaching
* Made by Manostat, New York, NY.


65
Fig
4-9
A corrosion cell in the flowing test


BLUE M CONVECTION
Fig. 4-10. Schematic of experimental configuration of continuous flow test.


67
vessel was kept at 90C in the tubing for some time (longer than 5
min), the temperature within the leaching vessel would not be changed
due to the water flow. The leachant after passage through the leach
vessel was collected weekly.
All the laboratory tests were run at 90C for up to 6 months.
The sample matrix of the experiments is shown in Table 4-8.
Before leaching, each sample was weighed and examined under an
optical microscope. FT-IRRS was run at 2-3 spots on each sample.
Analytical Techniques
A combination of several analytical techniques was used for
evaluating nuclear waste glass leaching. Each of the methods yields
averaged information which is characteristic of a volume extending
from the surface to a specific depth within the sample, as shown in
Fig. 4-11 and Table 4-9. In addition, SIMS provides depth resolved
concentration profiles from the surface into uncorroded bulk. As
discussed earlier, all the solid surface analysis techniques in Fig.
4-11 have been used for characterizing changes on the Stripa burial
glass surfaces. Solution analysis techniques including inductively-
coupled plasma (ICP), atomic absorption spectrophotometry,
colorimetry, and pH measurement were also used with the laboratory
leached specimens.
Solid State Analyses
Optical microscopy
Each sample was examined under a microscope using the reflection
light mode, both prior to and after leaching. Magnification of 100X


68
Table 4-8. Sample
Matrix of the Laboratory Tests.
Glass Composition*
SRL 1 65 + 29.8% TDS, Black Frit 165-Mobay
Temperature
90C
SA/V
0.1 and 1.0 cm 1
Leaching Time
1 3 and 6 months
Leaching Condition
Static and flow (0.1 and 0.3 mL/h), with and
without granite cup
Leachant
deionized water
Stripa ground water
Stripa ground water saturated with glass
powders for 14 days at 90C
*
Samples were run in duplicates.


69
RBS
iSil?T SEM-EDS
MICROSCOPE FT-IRRS SIMS/ION MILLING
5-200 A
0.5>¡m
I.Sjjm
BULK GLASS
SOLUTION ANALYSIS
ALTERED
LAYER
V
Fig. 4-11. Sampling depths with various techniques used in this
study (adapted from [63,64]).


70
Table 4-9. Characteristics of Analytical Techniques.
Sampling Spatial Detection
depth resolution Information limits($)
Secondary ion mass
spectroscopy (SIMS)
5-20A
(profiling
to =10pm
100A-=1pm
composition
structure
<1 0
Fourier transform
infrared reflection
spectroscopy (FT-IRRS)
=0.5 pm
3~5mm
composition
structure
morphology
3
Scanning electron
microscopy-energy
dispersive spectroscopy
(SEM-EDS)
1.5pm
1 5pm
morphology
composition
5
Rutherford back
scattering (RBS)
100 A~=1pm
1 mm
composition
>10~3


71
was used in all cases, which permits examination of the general
surface characteristics. For preleached glasses, heterogeneities,
such as crystallites, can be observed (Fig. 4-12). These
heterogeneities usually result when glass homogenization is not
complete or the wastes have not been dissolved by the glass matrix.
The glass surface finish conditions also can be checked (Fig.
4-13). In this study, glasses were polished to 600 grit or 6-pm
surface finish. Examination with an optical microscope served as a
quality control for the sample conditions. For the leached glass
surfaces, both surface roughening and surface precipitates can be
evaluated using this simple, rapid, and inexpensive technique.
Fourier transform infrared reflection spectroscopy (FT-IRRS)
Fourier transform infrared reflection spectroscopy (FT-IRRS) has
recently been developed as a semi-quantitative tool for
characterizing the surface structure and composition of glasses both
prior to and after exposure [65,66]. The important advantages of
this technique include (1) it does not require vacuum and energetic
electron or ion bombardment; thus it does not alter the surface of
the glass as may Auger electron spectroscopy (AES), electron
spectroscopy of chemical analysis (ESCA) and secondary ion mass
spectrometry (SIMS); (2) it is applicable to in-situ glass surfaces
of nearly any configuration and can be used for analysis of large or
small areas, if desired, is relatively inexpensive and requires only
standard infrared spectrometers; and (3) the FT-IRRS method can be
used as an automated analytical tool and can also be coupled with
solution analysis, making it especially suitable for characterization


72
Fig. 4-12. Light micrograph of SRL 131 + 29.8% TDS glass with
crystallites (100X).


73
Fig. 4-13.
Light micrograph of a typical glass surface
polishing to 600 grit surface finish (100X)
after


74
of surface/environment interactions [67-69]. In a spectrum of the
binary soda-silica glass surface, the region where the Si-O-Si
stretching peak (S) and silicon-oxygen-alkali (NS) stretching peak
overlap occurs is called the coupled region. Exposure of the glass
to a chemical environment alters the relative concentration of both
silica and alkali ions due to preferential leaching of the alkali
ions. This produces the decoupling of the S and NS peaks in the
infrared reflection spectra.
Extensive surface reactions can lead to roughening of the glass
surface due to formation of either pits or surface deposits.
However, the wavenumber location of the S and NS peaks is not changed
significantly by the surface roughening. Therefore, it is possible
to use the shift of the wavenumber location of the FT-IRRS peaks to
measure the change in composition of the glass surface, independent
of roughening or surface deposition. The extent of surface
roughening can be assessed by the decrease in intensity of the
FT-IRRS peak when wavenumber location remains unchanged.
It should be noted that the FT-IRRS technique is useful for
determining changes in reaction layers of -0.5 pm thick or greater.
Because of the 0.5-pm sampling depth, the information collected by
this technique within the sampling depth is averaged and accurate
analysis of very thin (<1000 A) surface corrosion films is not
possible. For very thick reaction layers, FT-IRRS provides an
analysis of only the outer -0.5 pm of the layer using near normal
specular reflectance. This nondestructive testing technique is
valuable for quick and efficient routine controls while in other


75
cases more detailed and usually more expensive analyses such as SIMS
are necessary. Figure 4-14 shows the FT-IRRS analyses of SRL 165 +
29.8? TDS glass/glass interface prior to and after 2-year burial in
Stripa at 90C. The decoupling of the S and NS peaks in the region
of 800-1150 cm-1 and loss of peak intensity as shown in the
postburial spectrum are a result of leaching.
Scanning electron microscopy/energy dispersive
spectroscopy (SEM-EDS)
The major advantages of SEM over other techniques such as
optical microscopy are that much higher magnification and a greater
depth of field are possible. The specimens were usually vacuum
coated with 100 A of C or Au-Pd. The information obtained using this
technique is mainly qualitative, although EDS in favorable cases may
yield the average composition of the outer most few microns. Figure
4-15 shows a typical SEM* micrograph of a glass surface after
polishing to 600 grit and prior to leaching. Figure 4-16 shows the
EDS data of SRL 131 + 2.98? TDS glass prior to burial.
Secondary ion mass spectroscopy (SIMS)
Secondary ion mass spectroscopy (SIMS) has an information depth
of the order of one atomic layer combined with ionic milling
(sputtering), which together with high detection sensitivity for most
elements offers a unique potential in profiling. During the last
several years, advances have been made by scientists at the Chalmers
* Scanning electron microscope, model JSM-35CF, JE0L Ltd., Tokyo,
Japan.
I


30
20 -
UJ
O
¡z

_J
Lx
L
cr
10
/ \
/ \
S/
1/ \
> \
I \
-Pre-burial
\
NS
\ ^ Post -burial
90c
\\
\\
\ \
\ \
\ \
\ .
\V
I
1200
1000 800
WAVENUMBER (cm"1)
600
Fig. 4-14. FT-IRRS analysis of SRL 165 + 29.8% TDS glass/glass interface prior to and after 2-year
burial in Stripa. The postburial spectrum shows the decoupling of the S and NS peaks in
the region of 800-1150 cm-1 and loss of peak intensity as a result of leaching.


Fig. 4-15. SEM micrograph of a typical glass surface after polishing
to 600 grit prior to leaching.


RELATIVE INTENSITY
KEV
Fig. 4-16. EDS analysis of an uncorroded SRL 131 + 29.8? TDS glass surface.


79
University of Technology, Sweden, to develop SIMS as a sensitive and
routine tool in the study of glass corrosion [70]. The glass samples
were coated with a 100 A Au film to reduce surface charging. The
Cameca 3-F ion probe accelerated and focused a beam of 0 ions
towards the glass sample, successively eroding the surface by
sputtering, while cyclically counting the yields of sputtered
secondary ions of different species which can be detected and
quantified with a mass spectrometer. The raw data, processed by an
on-line computer, consisted of these ionic yields vs the
corresponding sputtering time. With the aid of known relative
elemental sensitivity factors (RSF), the ionic yields were converted
to the percent atom concentrations of all the measured elements and
their sum of cations was set equal to 100 percent. Although H is
measured, it is not included in the conversion calculation, because
the H content of the preleached glass is unknown. The determination
of relative erosion speeds at different depths of the sputtered layer
permitted the conversion of sputtering time to depth. As an example,
the element concentration profiles of ABS 118 glass/glass interface
after 12-month, 90C burial are shown in Fig. 4-17.
In the most recent version, the profiles were corrected to
consider the elemental release and absorption during corrosion.
These profiles are different from the atomic concentrations (in
percent) shown in Fig. 4-17. The new profiles indicate the actual
gramatoms of each element after leaching of 100 gramatoms of
original glass, and so may be directly used in calculations of
elemental losses. Due to the cation (except H) release and


ATOMIC CONCENTRATION (%)
80
Fig. 4-17. SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 90C burial in Stripa. The atomic
concentrations of all cations (except H) are summed up to
100?.


81
adsorption, these actual gram^atoms of various cations may not be
summed up to 100 after leaching. These profiles were calculated
using the least leachable elements (in most cases Al, sometimes Fe,
Mn or Zr) as a standard and assuming that their actual gram*atoms
remain unchanged at any time at the surface.
Taking Al as a standard element, the following equation can be
wrriten based on 100 gram*atoms of unleached glass at a specified
depth in the glass surface,
GA.. . GA + GA GA... (4
Al,after Al,before Al,abs Al,leached
where GAftl after = gram*atoms of Al after leaching;
GAA1, before = Sram,atoms of Al before leaching;
GAAi at3S = granratoms of Al absorbed from solution;
GAA1, leached = Sramatoms of Al leached.
Since the concentrations of Al in the ground water were low both
before and after leaching (see Table 5-1), neglecting the last two
terms on the right side of equation (4-1) will not introduce
appreciable error. Thus, we have
GA.. , GA.. .
Al,after Al,before
(4-2)
where GAftl after and GAA1 before are same as in equation (4-1).
Also, we can write


82
GAA1,after at'^Al,after ^ ^i,after
(4-3)
where at.?A1 after is concentfation of A1 (in at.?) at a certain
depth after leaching as shown in Fig. 4-17 and E GA. is a
. i,at C6P
i
summation of the gram-atoms of element i after leaching. Combining
equations (4-2) and (4-3) gives
GA ^ at.? -2 GA.
Al,before Al,after l,after
i
(4-4)
Also
GA = at.?. -2 GA.
i.after l,after i,after
i
(4-5)
where at.?^ after is concentration of element i (in at.?) at a
certain depth after leaching, as shown in Fig. 4-17. From equations
(4-4) and (4-5), we have
at.?
GA.
i .after
GA
Al .before
i.after
at. ?
(4-6)
Al.after
Using equation (4-6), the actual gram-atoms of each element left
based on 100 gram-atoms of unleached glass at a certain depth can be
calculated. The results as obtained from the on-line computer of the
SIMS instrument are given in Fig. 4-18 for the same glass specimen
shown in Fig. 4-17.


GramatDms Remaining Based on
IOO Gramatoms of Unleached Glass
83
Fig. 4-18. SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 90C burial in Stripa. Data are presented
as gram*atoms of various cations remaining in the leach
layer at certain depth based on 100 gram*atoms of
unleached glass.


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NUCLEAR WASTE GLASS LEACHING
IN A SIMULATED GRANITE REPOSITORY
BY
BINGFU ZHU
A DISSERTATION PRESENTED TO THE GRADUATE SCHOOL
OF THE UNIVERSITY OF FLORIDA IN
PARTIAL FULFILLMENT OF THE REQUIREMENTS
FOR THE DEGREE OF DOCTOR OF PHILOSOPHY
UNIVERSITY OF FLORIDA

To my mother and late father

ACKNOWLEDGMENTS
The author acknowledges his gratitude to Dr. David E. Clark for
guidance and encouragement throughout his research. He also is
greatly indebted to Drs. Larry L. Hench, Lars Werme and George G.
Wicks for their advice and encouragement. The author thanks Drs.
Alexander Lodding, Christopher D. Batich, Stanley R. Bates and Gar B.
Hoflund for their timely suggestions, helpful discussions and review
of this dissertation.
The author wishes to thank Dr. Cheng Jijian for introducing him
to the field of chemical durability of glasses. Without his
guidance, the fulfillment of this research could not be possible. He
also thanks his wife, Jisi, for her support and encouragement.
iii

TABLE OF CONTENTS
Page
ACKNOWLEDGMENTS iii
LIST OF TABLES vi
LIST OF FIGURES viii
ABSTRACT xv
CHAPTERS
I INTRODUCTION 1
II PREVIOUS WORK ON NUCLEAR WASTE GLASS LEACHING 13
Laboratory Studies 13
General Considerations 13
Effect of Flow Rate 17
Surface Film Formation 20
Molecular Mechanism of Aqueous Dissolution 25
Systems Interaction Tests 29
Burial Studies 33
III RESEARCH OBJECTIVES, APPROACH AND SUMMARY OF
CONCLUSIONS 36
Research Objectives and Approach 36
Major Conclusions 37
IV MATERIALS AND METHODS 39
Glass Compositions and Characterization ...39
Burial Samples 39
Glass Quality 44
Laboratory Samples 48
Stripa Field Tests 48
Sample Assemblies, Minicans and Pineapple Slices...48
Stripa Repository 51
Burial and Retrieval 55
Disadvantage of the Burial Test Method 60
Similar Tests Being Used in MIIT Studies at WIPP...61
Laboratory Tests of Simulated Corrosion 62
IV

Analytical Techniques 67
Solid State Analyses 67
Solution Analyses 85
VTEST RESULTS 87
Field Test Results 87
General Observation 87
Results with ABS Glasses 89
Results with SRL Glasses 116
Effect of Glass Heterogeneities 125
Laboratory Test Results 135
Modified MCC-1 Static Leach Tests 135
Single-Pass Flow Tests and Static Tests Using
Rock Cups 140
VIDISCUSSION 152
ABS Glasses 152
SRL Glasses 1 61
A Model of Alkali Borosilicate Glass Leaching 168
Effect of Glass Composition 179
Influence of Repository Variables 185
Ground Water Chemistry 185
Effects of Repository Materials 1 88
Effect of Temperature 193
Comparison of Field and Laboratory Test Results 193
VIISUMMARY 197
REFERENCES 203
BIOGRAPHICAL SKETCH 212
v

LIST OF TABLES
Table Page
1-1 Quantity and Radioactivity of High-Level Nuclear Wastes
in Some Developed Countries 5
1-2 Candidate Waste Forms Considered for Geologic Disposal
of High-Level Waste 3
4-1 Nominal Waste Glass Compositions (wtí) Used in the
Stripa Burial 40
4-2 Sample Matrix of the Stripa Burial Tests 45
4-3 Variations in Spectral Characteristics of SRL Waste
Glasses 46
4-4 Nominal Composition of Black Frit 165-Mobay Glass 49
4-5 Average Major/Minor Chemical and Mineral Constituents in
Stripa Granite 54
4-6 Ground Water Composition and pH Measured in this Study
within the 1-month Test Hole at Stripa. Concentration
mg/L 58
4-7 Ground Water Composition for the Stripa Granite,
Literature Values 59
4-8 Sample Matrix of the Laboratory Tests 68
4-9 Characteristics of Analytical Techniques 70
5-1 Composition of Ground Water Collected from the Boreholes
where SRL Glass Pineapple Slice Assemblies Had Been
Buried 91
5-2 Gram*Atoms of Elements Remaining at Gel Mid-Plateau
and Outer Region of the Altered Glass Surface Based
on 100 Gram«Atoms of Unleached ABS 118 Glass after
12-Month, 90°C Burial in Stripa 119
vi

5-3 Relative Concentrations (Ratio to Si) at the Black Frit
165-Mobay Glass Surface after Static Leaching in the
Rock Cup Test. Data Are from EDS Analysis 147
6-1 90°C Glass Leach Rates During 12- to 31-Month Period
( pm/year) 155
6-2 SIMS Compositional Analysis of Glass/Glass, Glass/
Bentonite and Glass/Granite Interfaces for ABS 39,
ABS 41 and ABS 118 after 12-Month, 90°C Stripa Burial
(Gram*atoms Remaining Based on 100 Gram*atoms of
Unleached Glass) 158
6-3 Coordination Number and Bond Strength of Most Oxides in
Alkali Borosilicate Nuclear Waste Glasses 170
6-4 90°C ABS Glass Leach Rates During 7-12 Month Period 191
7-1 Estimated Boron Depletion .Depths (pm) after 300 Years of
the Thermal Period of Storage for the Six Nuclear Waste
Glasses 201
7-2 Estimated Boron Depletion Depths (pm) after 10'’
Years of Storage for the Glass/Glass Interfaces of SRL
Simulated Nuclear Waste Glasses 202
vi i

LIST OF FIGURES
1-1 Flow diagram showing the reprocessing of the spent
nuclear fuel 2
1-2 Schematic showing the glass waste form in a geological
repository 6
1-3 Glass structure containing dissolved wastes 11
2-1 Research activities on leaching of nuclear waste glass 14
2-2 Plot showing the total mass loss per unit area as a
function of flow rate 19
2-3 The five types of glass surfaces and six surface
conditions resulting from glass-environment interactions... 21
2-4 Ratio of normalized solubility to NLg^ (20 g/m^)
for CaC03, SrC03> Nd(0H)3, Fe(0H)3 and Zn(0H)2
in MCC-1 28-day test at 90°C in solutions of different
pH 26
2-5 The Si teachability of a borosilicate glass immersed
in a 5-day static 23°C solution buffered to various
pH values 27
2-6 Calculated Si concentrations in the surface layer and
bulk solution based on the surface layer diffusion
and pH as a function of leaching time. A diffusion
coefficient of 10 ° cm2/sec was assumed 30
4-1 Pineapple slices of glass, granite, stainless steel,
Ti, Pb and compacted bentonite before burial 43
4-2 Representative FT-IRRS spectra of SRL glass pineapple
slices before burial 47
4-3 A minican assembly 50
4-4 A typical pineapple slice assembly 52
viii

4-5 Seven preburial pineapple slice assemblies with different
sample stacking sequences for SRL simulated nuclear waste
glasses 53
4-6 Location within Stripa where SRL samples were buried 56
4-7 Diagram illustrating the position of the samples in the
Stripa mine during burial 57
4-8 Schematic of experimental configuration of static leach
test 63
4-9 A corrosion cell in the flowing test 65
4-10 Schematic of experimental configuration of continuous
flow test 66
4-11 Sampling depths with various techniques used in this
study 69
4-12 Light micrograph of SRL 131 + 29.8% TDS glass with
crystallites (100X) 72
4-13 Light micrograph of a typical glass surface after
polishing to 600 grit surface finish (100X) 73
4-14 FT-IRRS analysis of SRL 165 + 29.8% TDS glass/glass
interface prior to and after 2-year burial in Stripa 76
4-15 SEM micrograph of a typical glass surface after
polishing to 600 grit prior to leaching 77
4-16 EDS analysis of an uncorroded SRL 131 + 29.8% TDS glass
surface 78
4-17 SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 90°C burial in Stripa. The atomic
concentrations of all cations (except H) are summed up
to 100Í 80
4-18 SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 90°C burial in Stripa. Data are presented
as gram*atoms of various cations remaining in the leach
layer at certain depth based on 100 gram*atoms of
unleached glass 83
5-1 A typical assembly after burial in Stripa mine.
Bentonite coating can be observed on the outer surface
due to bentonite swelling 88
ix

5-2 Schematic of glass/glass interface illustrating several
types of surface areas resulting from water and/or
bentonite intrusion 90
5~3 FT-IRRS spectra of glass ABS 39 (a) and ABS 41 (b) before
and after 31-month, 90°C Stripa burial 92
5-4 SIMS depth profiles for (a) ABS 39 (Al-corrected) and
(b) ABS 41 (Si-corrected) after 31-month, 90°C Stripa
burial 94
5-5 Light micrographs (100X) of glass 39 (a) glass/glass,
(b) glass/granite and (c) glass/bentonite interfaces and
glass ABS 41 (d) glass/glass, (e) glass/granite and
(f) glass/bentonite interfaces after 31-month, 90°C
Stripa burial 97
5-6 SIMS depth profiles of boron for glass ABS 39 (a)
and glass ABS 41 (b) after 31-month, 90°C Stripa burial....99
5-7 FT-IRRS spectra of glass/glass, glass/granite and glass/
bentonite interfaces for nuclear waste glass ABS 118
buried in Stripa at 90°C for (a) 2 months and (b) 12
months. Also shown is the spectrum of a preburial
glass surface 100
5-8 Light micrographs (100X) of glass ABS 118 after
2-month burial, (a) glass/glass, (b) glass granite,
and (c) glass/bentonite interfaces, and after 12-month
burial, (d) glass/glass, (e) glass/granite, and (f)
glass/bentonite interfaces at 90°C in Stripa 102
5-9 SIMS depth compositional profiles of (a) B; (b) Cs,
Sr; and (c) Fe, U for ABS 118 glass/glass interface
after 2- and 12-month, 90°C burial in Stripa. The data
have been corrected using A1 concentration 103
5-10 SIMS depth compositional profiles of (a) Si, H, Na,
Li, K; (b) LD (including La, Ce, Pr, Nd and Y), P, Sn;
(c) Ca, Zn, Ba; and (d) Zr, Mo, Ni, Cr, Si for ABS 118
glass/glass interface after 12-month, 90°C burial in
Stripa 105
5-11 FT-IRRS analysis of ABS 118 glass/Pb, glass/Ti and
glass/Cu interfaces after (a) 2-month and (b) 12-month
90°C burial in Stripa 108
x

5-12 Light micrographs of ABS 118 glass surfaces after
90°C Stripa burial for 2 months, (a) glass/Pb, (b)
glass/Cu, and (c) glass/Ti interfaces, and for 12
months, (d) glass/Pb, (e) glass/Cu, and (f) glass/Ti
interfaces 109
5-13 Boron profiles of ABS 118 (a) glass/Pb, (b) glass/Cu
and (c) glass/Ti interfaces after 2- and 12-month, 90°C
burial at Stripa 111
5-14 Cs and Sr profiles of ABS 118 (a) glass/Pb, (b) glass/Cu
and (c) glass/Ti interfaces after 2- and 12-month, 90°C
burial at Stripa 112
5-15 Fe and U profiles of ABS 118 (a) glass/Pb, (b) glass/Cu
and (c) glass/Ti interfaces after 2- and 12-month, 90°C
burial at Stripa 113
5-16 SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90°C Stripa
burial, Si, H, Li, Na and K profiles 114
5-17 SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90°C Stripa
burial, Ca, Zn and Ba profiles 115
5-18 SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90°C Stripa
burial, LD, Pb, Cu and Ti profiles. LD stands for the
sum of La, Ce, Pr, Nd and Y 117
5-19 SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90°C Stripa
burial, Zr, Mo, Ni and Cr profiles 118
5-20 RBS analysis of ABS 118 glass/Pb, glass/Ti and
glass/Cu interfaces after 12-month, 90°C burial in
Stripa 120
5-21 FT-IRRS analysis of the glass/glass interface for three
SRL glasses after 2 years of burial in Stripa 121
5-22 SEM micrographs of glass surfaces in contact with glass
of the same composition during 2-year burial at 90°C
in Stripa: (a) SRL 131 + 29.8? TDS, (b) SRL 165 +
29.8? TDS, (c) SRL 131 + 35? TDS and (d) an uncorroded
glass surface 123
5-23 SIMS depth profiles of SRL glass surfaces after 2-year
Stripa burial at 90°C 124
xi

5-24 X-ray diffraction pattern for powders prepared from
devitrified SRL 131 + 29.8% TDS glass 126
5-25 SEM-EDS analysis of preburial SRL 131 + 29.8% TDS
glass: (a) homogeneous glass surface and (b) partially
devitrified glass surface 127
5-26 SEM analysis of SRL 131 + 29.8% TDS glass surfaces in
contact with bentonite, Stripa burial at 90°C: (a)
homogeneous glass, 1-month burial; (b) partially
devitrified glass, 1-month burial; (c) partially
devitrified glass, 3-month burial; and (d) partially
devitrified glass, 6-month burial 129
5-27 FT-IRRS analysis of SRL 131 + 29.8% TDS glass surfaces
in contact with bentonite, Stripa burial at 90°C: (a)
homogeneous glass, 1-month burial; and partially
devitrified glass (b) 1-month burial; (c) 3-month
burial; and (d) 6-month burial 131
5-28 EDS analysis of SRL 131 +29.8? TDS glass surfaces in
contact with bentonite, Stripa burial at 90°C: (a)
homogeneous glass, 1-month burial; and glass matrix of
partially devitrified glass, (b) 1-month burial; (c)
3-month burial; and (d) 6-month burial 133
5~29 EDS analysis of crystal areas of partially devitrified
SRL 131 + 29.8% TDS glass surfaces in contact with
bentonite, Stripa burial at 90°C: (a) for 1 month,
(b) for 3 months and (c) for 6 months 134
5-30 FT-IRRS analysis of SRL 165 + 29.8% TDS glass before and
after leaching for 28 days at 90°C in (a) deionized
water with SA/V = 0,1 cm-1, (b) Stripa ground water
with SA/V = 0.1 cm , (c) Stripa ground water with
SA/V = 1.0 cm 1 and (d) saturated Stripa ground water
with SA/V = 0.1 cm 1. Also shown is a spectrum for
the glass/glass interface after 3~month Stripa burial 136
5-31 SIMS analysis of SRL 165 + 29.8% TDS waste, laboratory-
corroded, 1 month 90°C in Stripa water with SA/V =
1 .0 cm'1 139
5-32 Solution pH vs time for Black Frit 165-Mobay glass
corroded under static and flow conditions 1 41
5-33 Concentrations of Si, B, A1 and Li in the single pass
flowing ground water at 0.3 ml/hr as a function of
leaching time for Black Frit 165-Mobay glass samples 142
Xll

5-34
Normalized leach rates of Li, B and Si as a function
of time under flowing (at 0.3 ml/hr) conditions for
Black Frit 165-Mobay glass with SA/V = 1.0 cm-1.
Also shown are the weight losses for glass samples
leached under static and flow conditions with SA/V
= 1 .0 cm"1 144
5-35 EDS analysis of Black Frit 165-Mobay glass leached in
the rock cup test at 90°C with SA/V = 1.0 cm 1 in
ground water under static conditions 145
5-36 FT-IRRS analysis of Black Frit 165-Mobay glass leached
in the granite rock cup test at 90°C under static
conditions with SA/V = 1 .0 cm 1 148
5-37 SIMS depth profiles for Frit 165-Mobay glass leached
in the granite rock cup tests at 90°C with SA/V =
1.0 cm , (a) under static conditions and (b)
under flow conditions (0.3 mL/hr) 150
6-1 Time dependence of reaction layer thickness for glass
ABS 39 (a) and ABS 41 (b) after 31-month, 90°C Stripa
burial 153
6-2 Boron depletion depth vs burial time for the glass/
glass, glass/granite and glass/bentonite interfaces.
Three ABS glasses are compared 1 60
6-3 Penetration depth as a function of leaching time for the
SRL glasses either buried in contact with glass,
stainless steel, granite or bentonite in Stripa mine, or
leached in Stripa ground water with SA/V = 0.1 or 1.0
cm 1 in laboratory 162
6-4 SIMS compositional profiles of SRL 165 + 29.8? TDS
glass/bentonite interface after 24-month, 90°C burial
in Stripa 165
6-5 Five modes of corrosion in partially devitrified
alkali borosilicate simulated nuclear waste glass:
(a) leaching of the glass matrix; (b) enhanced attack
of the glass-crystal interface; (c) pitting of the
polycrystalline phase at grain boundaries; (d) surface
films enriched in the less soluble multivalent
species; and (e) crystallite stripping 167
6-6 Stability of B2O0 and SÍO2 in aqueous solution
at 25°C as a function of pH 173
xiii

6-7 Schematics showing (a) the altered alkali borosilicate
glass surface and the compositional profiles after
leaching based on the model proposed in this
dissertation and (b) the altered glass surface
based on Grambow's model 176
6-8 The density index curve for three SRL glasses after
2-year burial in Stripa at 90°C 1 80
6-9 Compositional ternary diagram showing the direction of
increasing boron depletion depth. F^O represents
alkali metal oxide, represents A^O^ and
Fe2C>2, and WP stands for waste products 181
6-10 The boron depletion depth as a function of
(Si02 + A^O^/CF^O + 820^) wt ratio
in glasses 183
6-11 Schematic illustrating the relationship between
concentration, contact time and leach rate 186
6-12 The boron depletion depth as a function of burial time
for ABS 118 glass/glass, glass/granite, glass/Pb, glass/
Cu and glass/Ti interfaces after 90°C Stripa burial 190
6-13 SIMS compositional profiles of SRL 165 + 29.8% TDS
glass/glass interface after 8-10°C Stripa burial for
2 years 19^
xiv

Abstract of Dissertation Presented to the Graduate School
of the University of Florida in Partial Fulfillment of the
Requirements for the Degree of Doctor of Philosophy
NUCLEAR WASTE GLASS LEACHING
IN A SIMULATED GRANITE REPOSITORY
BY
BINGFU ZHU
May 1987
Chairman: Dr. David E. Clark
Cochairman: Dr. Larry L. Hench
Major Department: Materials Science and Engineering
Burial experiments of three Savannah River Laboratory (SRL) and
three Swedish alkali borosilicate (ABS) simulated nuclear waste
glasses were conducted to evaluate the resistance of these glasses to
ground water attack under repository-like conditions. Glass samples
were buried in the boreholes at a depth of about 350 meters below the
surface in the Stripa granite at either ambient mine temperature
(8-10°C) or 90°C. Included in the same boreholes were other
potential waste package components. Glasses were also leached in the
Stripa ground water contained in a leaching vessel under the
laboratory simulation conditions. The leached surfaces were
characterized using SEM-EDS, FT-IRRS, SIMS, RBS and optical
microscopy. Differences in glass leach rate were observed among the
six compositions with SRL 165 + 29.8% TDS being the lowest. Results
show that durabilities of the SRL composite nuclear waste glasses
XV

were increased by approximately six times when frit 131 was
substituted by frit 165. An increase of waste loading of SRL 131
glass from 29.8 wt% to 35 wtí decreases the teachability by a factor
of 2.
The leach rates of buried samples based on boron extraction at
90°C ranged from 0.3~3 pm/year for the glass/glass interfaces of all
glass formulations. These values are at least two orders of
magnitude lower than those for glasses leached using MCC-1 static
leaching procedures and deionized water. The Stripa repository-like
conditions can be simulated in the laboratory using Stripa ground
water and high SA/V ratios (_> 1.0 cm-1). Comparison of the
laboratory test results with field test results indicates that the
leaching mechanisms were similar under these test conditions. One of
the advantages of the laboratory simulation testing is saving of time
since glass leaches faster under the laboratory-controlled conditions
than under field-leach environment.
A model, based on glass structure and thermodynamic
considerations, is proposed to describe alkali borosilicate glass
leaching under repository-like conditions.
xvi

CHAPTER I
INTRODUCTION
The increasing use of nuclear energy for electric power
generation and the expanding application of radioisotopes in various
fields are inevitably associated with the production of growing
amounts of nuclear wastes. These wastes, which result from
fabrication, use and reprocessing of nuclear fuels, contain a variety
of hazardous materials. Hence, their disposal must ensure a low
probability of human contact.
Major types of nuclear wastes include high-level (HLW),
transuranic (TRU), low-level (LLW), uranium mine and mill tailings,
decontamination and decommissioning wastes, and gaseous effluents.
High-level wastes are usually further divided into those resulting
from either weapons production (defense waste) or commercial power
reactors.
Upon removal from the nuclear reactors, the depleted fuel is
stored under water for several months to permit the short-lived
fission products to decay. One of the options is to send the fuel
pellets to a chemical-reprocessing plant to recover the uranium and
plutonium, which are then available to make new fuel [1]. As shown
in Fig. 1-1, the reprocessing generally consists of dismantling
reactor fuel in a manner that permits dissolution of the core
material of the nuclear fuel pellets without dissolving their
1

Spent
fuel
clad
material
dismantling
->
core
material
solvent extraction
or ion exchange
U and Pu
recovered
Waste
Fig
1 -1 .
Flow diagram showing the reprocessing of the spent nuclear fuel [1].

3
corrosion resistant cladding [1], The resulting solution is
subsequently treated by several cycles of solvent extraction or ion
exchange to recover, separate and purify the residual uranium and
plutonium. At Savannah River Plant, Aiken, South Carolina; in
Hanford Reservation, Richland, Washington; and in Idaho National
Engineering Laboratory, outskirts of Idaho Falls, Idaho, there are
large facilities owned and operated by the United States government
that reprocess spent fuel coming out of the reactors used for making
weapons. However, only one commercial reprocessing plant, at West
Valley, New York, was ever operated in the U.S. Currently there is
no reprocessing of spent fuel coming out of commercial reactors in
the United States.
High-level nuclear wastes, whether reprocessed or not, contain
virtually all of the nonvolatile fission products, small amounts of
uranium and plutonium and all the other actinides formed by
transmutation of the uranium and plutonium in the reactors.
They can be generally characterized by their very intense,
penetrating radiation and their high heat-generation rates. The
fission products and actinides are the major concern since they
undergo spontaneous decay and emit radioactivity in the form of a and
1 37 90
8 particles, and Y-rays. The two elements, Cs and Sr , are of
most concern due to the relatively high concentration in the waste
and their decay time (i.e., - 30 years)[2] and concern for their
90
incorporation in body tissues, especially Sr in bones. When
decaying, they give off both heat and radioactivity for about 700
years [2], The actinides including U also emit radioactivity and

4
? vq
heat during decay—for example, about 25,000 years for Pu J , the
most abundant transuranium actinide [2].
Table 1-1 lists the quantity and radioactivity of high-level
nuclear wastes in some developed countries [3_5]. There are over
5 2
3X10" m" defense HLW stored at three government sites in the United
States. These defense wastes contain 1.6X10^ Curies of radioactivity
(1 Curie = 3-7X1010 disintegrations per second)[3]. There are over
2X1 O'* m^ commercial HLW containing about 1.1X101^ Curies of
radioactivity in the form of spent fuel in the United States [3].
The total amount of HLW in Europe, Japan, the United States and
U.S.S.R. is estimated to be 10.2X10'’ containing about 2.9X101^
Curies.
One method for disposal of HLW is immobilization in a high-
integrity solid waste form followed by emplacement in a mined cavern
at a suitable geologic repository [6]. As shown in Fig. 1-2, this
disposal system relies on multiple barriers to prevent the release of
radionuclides. The system includes
(1) solid waste form, a combination of host material (glass in
the illustrated case) and waste. The waste is incorporated
homogeneously in the host material to reduce the risk for
dispersion.
(2) a metal canister such as stainless steel, which is welded to
form a hermetically sealed container after the waste form is
placed in it.
(3) a metallic overpack, of such materials as e.g., mild steel,
ductile iron, pure titanium, or titanium alloy (Ti Code-12),

5
Table 1-1. Quantity and Radioactivity of High-Level Nuclear Wastes
in Some Developed Countries.
Radioactivity Quantity
Form (Ci) (nr) Source
U.S. Defense
slurry
sludge
1.6X109
3X105
[3]
U.S. Commercial
spent
fuel,
sludge (West
Valley, NY)
1 .1 XI01 0
2X1 0^
[3,4]
Europe Commercial
spent
fuel
-
2X1 05
estimated
Japan Commercial
spent
fuel
-
7X1 04
estimated
U.S.S.R.
slurry
sludge
-
2.5X10^
estimated
Total
2.9X1010
10.2X105
[5 ]

6
m
Fig. 1-2. Schematic showing the glass waste form in a geological
repository.

7
and nickel alloys [7], which serves as an additional barrier
for radionuclide containment.
(4) a sleeve, when required, which is used to assure clearance
for the retrievable package to facilitate its removal during
the retrieval period. It provides structural support
against geologic pressure forces and may also serve as a
barrier for radionuclide containment.
(5) backfill, the material contained between the other
engineered waste package components and the host rock, which
serves to facilitate heat transfer, load transfer and
compatibility of the other engineered waste package
components with the host rock. It may also serve as one of
the barriers for radionuclide containment and a sorptive
medium for radionuclide release. Swelling clays such as
bentonite, alone or in a mixture with quartz or other
minerals, are being considered as backfill materials.
(6) a buffer, the material used to facilitate conditioning of
the ground water, immobilization of radionuclides and
compatibility of materials.
(7) a filler, which is any material used to fill space between
other components of the engineered waste package and may or
may not have other specified functions.
Five years ago, there were seven candidate waste forms chosen
for geologic disposal of HLW in the United States (Table 1-2). After
a multifaceted assessment [8-11], borosilicate glass and Synroc (a
titanate-based polyphase crystalline ceramic material) were selected

8
Table 1-2. Candidate Waste Forms Considered for Geologic Disposal of
High Level Waste [8].
Waste Form
Comments
Borosilicate Glass
Primary Waste Form, U.S. Reference Waste
Form
Synroc-C,D
Alternative U.S. Waste Form
Tailored Ceramic
Semi-finalist U.S. Alternative Waste Form
High-Silica Glass
Semi-finalist U.S. Alternative Waste Form
FUETAP Concrete
Semi-finalist U.S. Alternative Waste Form
Coated Sol-Gel Particles
Semi-finalist U.S. Alternative Waste Form
Glass Marbles in Lead
Matrix
Semi-finalist U.S. Alternative Waste Form

9
from the seven as the primary waste form and first alternative,
respectively. The focus of this work is on borosilicate nuclear
waste glass.
There are two major reasons why glass was selected as the
primary waste form. First, any material used for encapsulating
radioactive wastes must be capable of surviving for at least 10,000
years in a wide range of severe environments. Glasses can meet this
requirement. The existence of natural glasses, such as obsidians,
basalts, or tektites, which are millions of years old, demonstrates
that glass can be formulated which will survive long-term
environmental exposures. Similarly, synthetic glasses of known
longevity or performance, such as Roman glasses buried in the
Mediterranean or exposed to ground water for nearly 2,000 years, also
demonstrate the potential long-term performance of nuclear waste
glass. Second, the process for producing nuclear waste glass is
fairly simple. It involves feeding a slurry of waste sludge and
glass frit to a continuous glass melter, from which the molten waste
glass is poured into a canister. Such a simple fabrication method
makes the remote control of the whole process possible, as
demonstrated in the United States and France in full-scale operations
[2,12].
In contrast, consolidation and synthesis of the mineral phases
in synroc require hot isostatic processing or uniaxial hot
processing, which complicates the remote production processes.
Although the uranium leach rates are higher and the waste loading is
lower for the glass form than for the crystalline ceramics,

10
borosilicate glass is currently the choice of most countries as the
primary waste form due to simplicity of fabrication, moderate waste
loading, intermediate product performance and radiation stability.
The list of candidate sites for the first repository in the
United States has been narrowed to three locations—one in Nevada in
volcanic tuff, one in Texas in salt, and one in Washington state in
basalt [2]. Other rock formations such as granite in Sweden have
been considered outside the United States [13]. A final decision on
the site in the United States is still several years away and will
require extensive testing and risk assessment.
The major concern when the waste is buried deep in the ground is
that it might come into contact with water and be transported back to
the earth's surface. Therefore, the resistance of the solid waste
form to underground water attack is a problem of major concern,
because the second innermost barrier (canister materials) is only
expected to survive about 1,000 years in a geologic environment [14].
A nuclear waste glass is defined as a single phase amorphous
material in which quantities of both radioactive and nonradioactive
oxides are dissolved. The concept of using glass as a host for
radioactive waste is based upon the radionuclides entering into and
becoming part of the random three-dimensional glass network. Figure
1-3 schematically illustrates a portion of an alkali borosilicate
glass network containing various radionuclides as constituents. The
4-
structural network of the glass is provided primarily by [SiO^] ,
5- 3-
[BOjj] and [BO^] polyhedra. Neighboring polyhedra are bonded
together by sharing strong ionic-covalent bridging oxygen bonds.

11
9
OXYGEN
SILICON
BORON
Fig. 1-3. Glass structure containing dissolved wastes (adapted from
[15]).

12
+ 2 + 2
Other multivalent species such as Fe ’ , rare earths or actinides
are also generally bonded within the network by bridging oxygen
bonds. Low valence ions, such as Na+, Cs+, Sr+2, etc., are bonded
into the network by sharing various nonbridging oxygen bonds,
depending upon size and valence of the ions. This difference in type
of bonding in the glass network is responsible for the complex leach
behavior of nuclear waste glasses.
To date, there are only few data available regarding the
leaching behavior of nuclear waste glasses in the presence of a
variety of disposal system components [16-23]. In order to test
possible synergistic interactions of the materials in a nuclear waste
disposal system under repository-like conditions, in situ burial
experiments were designed. Such experiments approximated the
physical conditions of the repository more closely than laboratory
tests. Laboratory systems tests were also designed, when necessary,
to evaluate the effects of individual system variables on glass
leaching performance.
The primary objective of this dissertation was to determine the
leaching performance of the glass containing high-level nuclear
wastes* under a simulated repository condition and to investigate how
this is affected by the presence of other waste package components
and geologic conditions. In the process of achieving this goal a new
model of glass leaching was developed that satisfactorily describes
the observed results from both laboratory and field studies.
* The wastes used in this dissertation were simulated. It is assumed
that isotopes of the same element have similar chemical behavior.

CHAPTER II
PREVIOUS WORK ON NUCLEAR WASTE GLASS LEACHING
Extensive laboratory tests and some field tests have been
conducted using various combinations of reference materials in order
to evaluate their effects on glass leaching. Figure 2-1 summarizes
the research activities on nuclear waste glass leaching
performances. The laboratory tests performed include static and flow
experiments. In most of the laboratory tests, deionized waster was
used. Glasses were also leached in synthetic ground water, such as
silicate water and brine, and/or in the presence of other waste
package components. Relatively limited burial tests include a
15-year burial at Chalk River, Canada [24,25], a 9-year burial in
England [26] and more recently an initiated burial study in Belgium
[22]. More extensive work has been carried out in the Stripa mine in
Sweden [16-20]. The major focus of this investigation is on the
Stripa burial and laboratory systems interactions. The Waste
Isolation Pilot Plant (WIPP) program was designed based on the
experience from the Stripa burial test. This is the first burial
test to be conducted in the United States.
Laboratory Studies
General Considerations
For some time, the primary issue of concern regarding glass and
other HLW forms has been long-term stability in contact with hot
13

Fig
2-1
Research activities on leaching of nuclear waste glass.

15
repository ground waters in the event a canister is breached. In the
early 1980s, five tests were developed to determine the chemical
durability of waste forms [27,28] under either static (MCC-1P* and
MCC-2P) or flowing (MCC-4S and MCC-5S) leaching environments.
Maximum release by waste forms is determined using powders and
stirred solutions (MCC-3S). General acceptance of these test
methods, initiated by the Materials Characterization Center [27,28],
reduced inconsistency, improved communication and made possible the
comparison of data collected from different laboratories. This
facilitated the accumulation of an extensive data base on glass
leaching, including nuclear waste glasses.
In this paper, the term "leaching" is defined as release of
glass component oxides or elements through glass-aqueous solution
reactions without regard to mechanisms of release. The term
"corrosion" is also associated with deterioration of glass surfaces
due to the reactions that occur when water interacts with glass.
Therefore, these terms are used synonymously in this dissertation.
Most leach test data are reported for short periods of time,
i.e., 28 days or less. Such short-term data are frequently used to
compare the relative stability of waste forms and to study effects of
variables that control the rate of leaching. For example, Plodinec
et al. [29] used the approach of Newton and Paul [30] to predict
* Materials Characterization Center, Pacific Northwest Laboratory,
Richland, WA.

16
nuclear waste glass leaching based on thermodynamic aspects of its
chemical composition. They found a linear relationship between log
normalized mass loss of Si (g/m , 28 days) and free energy of
hydration (kcal/mol) for a number of natural and synthetic glasses,
including simulated nuclear waste glasses. Comparing the corrosion
resistance of nuclear waste glasses to natural glasses and ancient
man-made glasses and/or relative thermodynamic stabilities allows
extrapolation of waste glass corrosion resistance to geologic times
[31,32].
Strachan [33] has reported results from a 1-year-leach test
using MCC-1 static test procedures. He found that a dramatic
decrease in the rate of leaching occurred after approximately 91
days. The PNL76-68* glass appeared to continue to alter, albeit at a
significantly reduced rate, even though the solution concentrations
of many elements were saturated or supersaturated with respect to
alteration phases. In his studies, glasses were leached in deionized
water, silicate water and brine at either 40°, 70° or 90°C with the
ratio of glass surface area to volume of leachout (SA/V) of 0.1
cm 1. Solid state analyses of the leached specimens indicated a
steady growth of two layers. The outer layer grown by precipitation
reactions on the original surface of the glass consisted
predominantly of zinc and silicon, thus indicating a zinc silicate
phase(s). An altered layer remained behind as the aqueous solution
* A nuclear waste glass composition developed by Pacific Northwest
Laboratories, Richland, WA.

17
leached soluble and moderately soluble material from the glass
matrix. Thus, this layer was rich in Fe, Nd, La, Ti, and depleted in
B, Cs, Na and Mo. The altered layer thicknesses for specimens
leached in deionized water, silicate water, and brine at 90°C for up
to 1 year ranged from 30 to 50 pm.
The long-term data obtained by Bates et al. [3*0 using deionized
water and MCC-1 test conditions agree qualitatively with those
obtained by Strachan [33]. However, data of Bates et al. [34]
indicate that the normalized elemental mass losses for most elements
are constant after approximately 6 months of leaching, whereas
Strachan's data indicate that leaching losses are continuing for
several elements at the end of one year. The following equation
2
defines the normalized elemental mass loss NL.^ in g/m :
NL.
i
(2-1 )
where m^ = mass of element i in the leachate, g;
f^ = mass fraction of element i in the unleached specimen,
dimensionless;
p
SA = specimen surface area, m .
One important feature observed in results of Strachen [33] and Bates
et al. [34] is the preferential leaching of B and Na. The normalized
elemental mass losses for these two elements are larger than for Si.
Effect of Flow Rate
It is recognized that under certain conditions ground water will
flow through a geological repository and react with its contents.

18
Strachan et al. [35] have reported increased leach rates for Si and
Sr at a flow rate of 6 mL/h compared to static testing. Similar
results have been found by other workers [36]. Based on weighing the
samples before and after corrosion, the rate of leaching increased as
the flow rate was increased from 0.1 to 10 mL/h. Little difference
was observed between the static test and the test in which the flow
rate was 0.1 mL/h during the first 28 days.
It has been found [36] that at sufficiently low flow rates
(between 0.1 and 2 mL/h in Fig. 2-2) the total mass loss per unit
area of waste glass matrix components surface is directly
proportional to flow rate. The concentration of glass components in
the leachate is nearly independent of flow rate. In the case of some
matrix components (Si, Al), this concentration is determined by
saturation with respect to the surface of the glass, as modified by
the leaching process and possible alteration reactions. The modified
surface forms a barrier against migration.
At sufficiently high flow rates (> 2 mL/h in Fig. 2-2), the
release rate becomes constant, limited by the kinetics of the
leaching processes. In this case, corrosion products as well as
potential surface-passivating species are removed from the leaching
vessel, reducing the beneficial effects on both solution saturation
and protective surface film formation.
Present indications are that high flow conditions (>10 mL/h) are
very unlikely in a geologic repository [37]. Low flow conditions are
expected in the repository and the leach rate of the glass will be

19
¡non.
CM
O'
LiJ
2:
b
0Í
id
D.
(/)
CO
o
CO
co
<
t
o
R-
u
SRL ¡31-29.3% TD3-3A
90° C, D.!. Wafer
SA/ V = 0.! cm-
Basalt ¡
tuff
granite ! Potential Flow Rates in Repositories
-l
suit
!
I
io r
0
0
J
G.i(mL/hr) C).3(mL/hr) l.p(rnL/yr) IC'mL./S'ir}
46(nt/yr) 23(m/yr) 46(m/yr) 464(.n/yr)
LOG
LOW RATE
Figi 2-2. Plot showing the total mass loss per unit area as a
function of flow rate (adapted from [36]).

20
limited by the rate of transport of corrosion products from the
repository.
Surface Film Formation
Previous efforts to generalize the surface behavior of silicate
glasses proposed five types of glass surfaces and six surface
conditions to represent a broad range of glass-environment
interactions [38-40]. The type of surface is dependent on the
environmental history of the glass and may be defined in terms of
surface compositional profiles, as shown in Fig. 2-3. The ordinate
in Fig. 2~3 represents the relative concentration of SÍO2 (or oxides
in Type IIIB surface) in the glass and the abscissa corresponds to
the depth into the glass surface. If species are selectively
dissolved from the glass surface, the relative Si02 concentration
will increase producing a Si02-rich surface layer. If all species in
the glass are dissolved simultaneously (congruent dissolution), the
relative concentration of Si02 will remain the same as in the
original glass. When combinations of selective dissolution,
congruent dissolution and precipitation from solution occur, then any
one of the six surface conditions shown in Fig. 2-3 is possible.
Hench [40] has pointed out earlier that low leach rates of
complex nuclear waste glasses are due to Type IIIB surfaces, which
are composed of multiple layers of oxides, hydroxides and hydrated
silicates resulting from a sequence of solution-precipitation
reactions between the glass surface and leaching solutions (Fig.
2-3). A number of alkali borosilicate nuclear waste glasses that
exhibit Type IIIB surface behavior have elemental leach rates as low

21
CM
O
TYPE I
Original glass solution interface
\ ! BULK
z
o
o
CO
Cv'
O
CO
Inert glass
TYPE II
'Selective leaching
z
o
»-
3
á
to
-BUU<-
Pro»«ctiv« film
on glass
-OISTANCE-
-DI STANCE-
TYPE III B
Fig. 2-3. The five types of glass surfaces and six surface
conditions resulting from glass-environment interactions
(adapted from [40]).

22
2
as 0.02 to 0.2 g/m .day with a time dependence of static leaching of
to t0'^ or less after 28 days at 90°C.
In this work terms such as selective leaching and congruent
dissolution are used in discussing the glass leaching mechanisms.
Selective leaching includes ion exchange of the mobile species in the
glass and selective dissolution of glass matrix, structural or
network species with or without precipitation. Ion exchange involves
a process in which exchange between mobile species such as Na from
the glass and hydrogen or hydronium ions from the solution occurs.
Ion exchange can also occur between Ca, Mg and K in ground water with
mobile species from the glass. During this process, the remaining
constituents of the glass are not altered. As mentioned in Chapter
I, the structural network of the borosilicate glass is provided by
[SiOij]14 , [B0¿|]^ and [BO^]^ polyhedra. Since different glass
network formers dissolve at different rates, selective dissolution of
matrix, structural or network species is usually observed with a
multicomponent glass containing two or more network formers. This
may or may not be followed by precipitation depending on the
composition of glass and solution. Congruent dissolution occurs when
the species comprising the glass are dissolving into solution in the
same ratios as they occur in the bulk glass. Without precipitation
the composition in the glass surface is not changed by congruent
dissolution. However, large dimensional changes often accompany such
kinds of corrosion. Congruent dissolution may be followed by
precipitation after certain less soluble species approach
saturation. In this case, the composition of the glass surface

23
changes away from that of the bulk glass and less soluble
constituents are enriched at the altered layer.
Although a short (several days) period of predominant alkali-
hydrogen ion exchange may occur for Type IIIB glasses, the dominant,
long-term mechanism controlling corrosion is a combination of more or
less selective dissolution of glass matrix followed by
precipitation. The extent of matrix dissolution and onset of surface
and inner precipitation will depend on the time required for various
species in the glass to reach saturation in solution. Saturation of
a certain species will be a function of the initial solution pH,
concentration of alkali species in the glass and their rates of
release which change the solution pH, temperature, initial
concentration of that species in the solution, the ratio of glass
surface area to volume of leachant (SA/V) which influences solution
concentration, and flow rate which also affects solution
concentration.
The theoretical basis for Type IIIB glasses is the investigation
of Grambow which predicts the formation of a series of insoluble
reaction products on glass surfaces [41]. He concluded that reaction
of the matrix is the fundamental process that occurs in the leaching
of alkali borosilicate nuclear waste glasses. He pointed out that
without solubility restrictions, congruent dissolution occurs at all
pH values and leachant compositions. That is, the glass dissolves
congruently at a rate proportional to kt1. Even after saturation has
occurred with respect to a certain species, the glass can continue to
dissolve congruently with simultaneous precipitation of that species.

24
When solution saturation of species "i" is reached, there is no
longer any driving force for that species to leave the glass
surface. Consequently species i will accumulate at the glass-
solution interface as the matrix dissolves. If matrix dissolution
releases alkali ions, as will be the case for most glasses, there
will be a concomitant rise in pH proportional to the flow rate or
SA/V of the system. An increase in pH can have several simultaneous
effects on the glass, the solution and the glass-solution
interface. At the new pH, a second species "j" may reach solution
saturation and subsequently be retained in the glass surface along
with species "i." The extent of apparent incongruent dissolution of
the glass is thereby increased. The sequence of events that occurs
is predictable, based on the solubility limits of each species at a
given pH, as shown by Grambow [41].
Figure 2-4 summarizes the behavior of various elements
considered by Grambow [41]. Here, the ratio NS^/NLg^ is used to
present the solubility limits of these elements in solutions of
different pH. The normalized solubility NSi (g/m ) is given by
NS.
i
C
f.
i
i, sat
• SA/V
(2-2)
where ^ = the solubility limited concentration in the leachate
at the specified conditions, g/L;
f^ = the mass fraction of element i in the glass;
2
SA = the specimen surface area, m ;
V = the volume of the leachage, L;

25
p
NLg^ = the normalized elemental mass loss of Si, g/m , as
defined in equation (2-1).
Therefore, in static leach tests, nuclear waste glass containing Fe
oxides should concentrate Fe within surface layers. Zinc, Nd, Sr and
Ca should be concentrated as well in nearly neutral or slightly
alkaline solutions with Na and B depleted.
The low overall leachability of many nuclear waste glasses over
a pH range from 4.5 to 9.5 is a consequence of the formation of the
multiple barrier (Type IIIB) films. Figure 2-5 is a plot of Si
leachability for an SRL composite waste glass immersed in a 5~day
static 23°C solution buffered to various pH values from 3-5 to 10.7
[42]. These data show that over the pH range expected for repository
ground waters, indicated by arrows, glass leachability is low, since
the formation of less soluble reaction products lowers the solubility
of silica in the solution [43].
One thing that Grambow did not explain with his leach data is
why NL^/NLg^, where i is Ca, Fe, Zn, Nd or Ce, is usually larger than
1 when the solubility restrictions are removed. If NL^/iNLg^+NLg)
had been used, it may be much closer to 1, since BgO^ is also a glass
network former.
Molecular Mechanism of Aqueous Dissolution
In discussing the molecular mechanism of aqueous dissolution of
alkali borosilicate glasses, Grambow [44] extended the idea of
Aagaard and Helgeson [45] on the pure silica-water interactions, and
interpreted the observed saturation effects as a local surface
equilibrium process involving the critical activated surface

26
Fig. 2-4. Ratio of normalized solubility to NLg• (20 g/m2) for
CaC03> SrCOn, Nd(0HK, Fe(0H)o and Zn(0H)2 in MCC-1 28-day
test at 90°C in solutions of different pH (adapted from
[41]).

Leachabiüty, g/m
27
6
pH
Fig. 2-5. The Si leachabiüty of a borosilicate glass immersed in a
5-day static 23°C solution buffered to various pH values
(adapted from [M2]).

28
complex: for every complex desorbed from the glass matrix, another
complex is adsorbed from solution (equal forward and back
reactions). However, compared to the silica-water system, the waste
glass-ground water system is much more complex; dozens of elements
are involved in the system in addition to a dependence on variables
such as Eh, pH, ground water composition. Furthermore, when using
the approach of Aagaard and Helgeson, it is assumed that rates of
hydrolysis are controlled primarily by reaction kinetics at activated
sites on the surface of glass and not by diffusional transfer of
material through a leached outer zone or a coherent surface layer of
reaction products. Grambow [44] assumes that there exists a critical
surface complex whose desorption controls the mobilization of various
glass constituents. In contrast to saturation of Ca, Fe, Nd, and
others, saturation of silica in solution has a major effect on the
corrosion rate. Silica is the dominant constituent of the activated
complex and, according to Grambow's arguments, its desorption (as
silicic acid) from the glass network will limit the rate of release
of other glass constituents, even when these elements are not
solubility limited in solution. At saturation, condensation of
silanol groups will stabilize the glass network against further
attack of aqueous species.
Recent data from flow tests [46] as well as other investigations
[47] indicate the importance of considering diffusion in the leached
layer in addition to the reaction kinetics at the activated sites on
the surface of glass. Data from flow tests [46] indicate an initial
increase to a maximum of the solution concentration of various glass

29
constituents followed by a decrease with time. Grambow et al. [47J
speculate that the leaching is controlled by the transport of silicic
acid through a growing surface layer, as shown in Fig. 2-6. In this
figure the saturation concentration is not a constant because the pH
in the surface layer varies with time. Surface layer diffusion
results from the difference between the silicic acid concentration in
the bulk solution and at the surface layer. In the surface layer,
saturation is reached after 200 days, whereas the solution
concentration is still far below saturation. The same general trends
are observed for other elements such as Na and B [46].
Systems Interaction Tests
A comprehensive systems interaction study was performed by Wicks
et al. [48] in which they compared leaching behavior of a defense
waste glass in deionized water, ground water and both waters
containing rocks. Their analyses were based on the concentrations of
species in solution and did not take into account the species
adsorbed onto solids present in the leaching vessel. They found that
the presence of salt (from Carlsbad and Avery Island), basalt, shale,
granite and tuff all slightly decreased the concentrations of glass
species in solution compared to those obtained when deionized water
was used alone. Similar results were found with synthetic ground
waters and with the same water containing the various rocks. Actual
ground water yielded results comparable to the MCC reference waters
and synthetic ground waters.
Clark and Maurer [49] have investigated the effects of several
types of rocks, including basalt and granite, on the leaching of a

Si CONCENTRATION (mg/I)
30
_r0ZZr.ZZZ=nmnzr——_ . 1
0 100 200 300 400
TIME (d)
Fig. 2-6. Calculated silicon concentrations in the surface layer and
bulk solution based on the surface layer diffusion and pH
as a function of leaching time. A diffusion coefficient
of 10 9 cm2/sec was assumed [46,47].

31
borosilicate glass. With the possible exception of granite, the
combination of glass and rocks in the same leaching vessel did not
appear to have any significant effects in a 28-day test.
Brine solution generally decreases the rate of glass corrosion
[48,50,51], with the possible exception of Sr, Ce and similar
elements [35]. In the brine solutions, a protective magnesium
chloride complex forms on the glass surface. Exposure of nuclear
waste glass to tuffs results in a small decrease in corrosion rate,
perhaps due to a buffering effect [48]. Autoclave tests of basalt-
glass interactions [52] and granodiorite-glass [53] interactions show
a decreasing rate of attack.
McVay and Buckwalter [54] investigated the effect of iron on
nuclear waste glass leaching. They found that the presence of
ductile iron in deionized, tuff and basalt ground waters containing
PNL 76-68 borosilicate glass caused significant changes in the
leaching characteristics of the glass. Formation of iron silicate
precipitates effectively removes many elements from solution and
therefore inhibits the saturation effects which normally cause
decreases in elemental removal rate. Thus, basalt and tuff ground
waters behave similarly to deionized water in the presence of ductile
iron. The precipitates also retard saturation effects, resulting in
high sustained leach rates and thus greater total elemental removal
from the glass. A synergistic effect occurs between the two
materials. The iron enhances glass leaching and the glass enhances
iron corrosion.

32
The presence of a radiation field during storage and its effect
on glass leaching is another consideration. Radiation may affect the
glass-water system in several ways. Gamma radiation has been found
to result in approximately three- to seven-fold increases in the
leach rates of borosilicate glasses [55,56]. As reported by McVay
and Pederson [55], some of the enhancement is due to nitric acid
formation from air radiolysis in the presence of water. Nitric acid
appears to preferentially attack zinc and lanthanides, both of which
normally build up on the surface of the PNL 76-68 glass when leached
in nonacidic solutions. The change of the solution chemistry by
gamma radiation and generation of reactive species such as OH- from
water radiolysis also appear to be important. The principal effect
of water radiolysis products is the increased silicate dissolution.
The leaching behavior of the radioactive glass has been
investigated in comparison to that of the simulated glass [57]. In
this case it was found that radiation, due to the low dose rate with
the radionuclides (0.594 Ci per specimen), does not affect
significantly the leaching rate. This conclusion includes the
effects of radiation damage to the glass itself and the interaction
of the radiation field from the glass with the water and air.
Profiles of Pu and U behave similarly during leaching, both being
enriched in the surface of the glass. Leaching of radioactive glass
results in loss of B, Na, Li and Mo with about the same depth of
leaching. The leaching mechanisms appear to be similar for
radioactive and nonradioactive glasses [57].

33
Burial Studies
Burial studies were started in the late 1950s and early 1960s.
Merritt and Parsons [24,25] pioneered two tests of high-level waste
(containing real radionuclides) incorporated into nepheline syenite
glass and buried in contact with ground water for 15 years at Chalk
River, Canada, at ambient temperature. Fletcher [26] conducted
burial experiments of waste glass samples in England for up to 9
years. Although the field tests were not performed under actual
repository conditions, they did provide an approximation to a
potential repository. Preliminary results from burial experiments
[24,25] have shown that glasses leached at much lower rates under
repository-like conditions than under laboratory conditions. As an
example, the observed field leach rate from the Canadian burials was
over 200 times lower than the lowest leach rate reported in the
laboratory [24,25]. The authors attributed about 1/5 of this
difference to the lower aggressiveness of ground water over distilled
water used in the laboratory experiments and to its lower temperature
(6°C in the field compared to 25°C in the laboratory). The remainder
of the difference was attributed to the formation of a protective
surface layer.
The leaching performance of a waste glass depends on the
environment under which it is tested. In a repository, the system
variables ultimately controlling the environment to which the waste
glass is exposed include geology, engineered waste package
components, initial ground water chemistry, temperature, pressure,

34
radiation field, water contact time and flow rate through the
repository.
The most extensive and systematic field tests began in 1982 and
involve deep burial (350 meters below surface) in granite in the
Stripa mine, Sweden [16-20]. The boron depletion depths of glass
ABS 39* and 41* ranged from 0.2 pm to 15 pm, depending on composition
and the type of material to which the glass was exposed after 1 year
of burial at 90°C. At the glass/glass interface, both glasses showed
a depletion of Na, Cs and B, but for the more corrosion-resistant
glass, the lower depletion depth was ascribed to the formation of a
thin (0.2 pm) coherent and dense outer layer, enriched in Mg, Ca, Sr,
Ba, Zn, Al, Fe and Si, which impedes both the ion exchange and
network attack of the bulk glass underneath. The presence of
bentonite increased the boron depletion depth up to 1 year by a
factor of approximately 5, whereas granite decreased this depth by
about 2 times. This behavior is attributed to bentonite serving as a
semi-infinite ion exchange medium where Ca from the bentonite is
replacing Na, Li and B from the glass [19]. In contrast, the small
congruent solubility of granite seems to augment the glass in
reaching solubility-limited leaching [21].
Another in situ test was initiated in 1986 and involves burial
in a clay formation in Mol, Belgium [22]. A number of simulated
waste forms (including HLW glasses and glass-ceramics) have been, or
will be, buried at the site. Their corrosion rate will be measured
* Swedish alkali borosilicate (ABS) and nuclear waste glasses.

35
in two environments susceptible to contact the radioactive waste
during its geological storage in a clay formation: host clay and a
humid atmosphere loaded with clay extracts. The tests, with total
exposure times of 6 years, will be carried out at various
temperatures, 15, 50, 90 and 170°C.

CHAPTER III
RESEARCH OBJECTIVES, APPROACH AND SUMMARY OF CONCLUSIONS
Research Objectives and Approach
The primary objectives of this study were (1) to evaluate the
leaching behavior of selected nuclear waste glasses in a realistic
repository environment, (2) to develop a characterization methodology
for comparing field data with laboratory data and (3) to assess
leaching mechanisms and to correlate the mechanisms observed in
laboratory-leached vs field-leached specimens.
In order to achieve these objectives, both field experiments and
laboratory simulation tests were conducted. The field tests involved
long-term (up to 31 months) deep burial (350 m below surface) in
granite in the Stripa mine in Sweden. Two configurations of samples
were used. One was a 32 mm in diam. x 35-mm long minican where an
alkali borosilicate glass with simulated HLW was cast into stainless
steel. The second configuration was the so-called "pineapple
slices," 51-mm in diam. x 5-mm thick, which resulted in a variety of
glass/repository materials interfaces. Two temperatures, 90°C and
the ambient temperature (8-10°C), were used to simulate the
repository conditions during and after the thermal period of storage,
respectively. Comparisons were made of six alkali borosilicate
simulated nuclear waste glasses, including three American Savannah
River Laboratory (SRL) glasses and three Swedish alkali borosilicate
36

37
(A3S) glasses. Different glass/repository materials interfaces were
provided to investigate effects of these materials on glass leaching.
In the laboratory, methods were designed to consist of both
static and single-pass low-flow tests, using granite rock cups as
leach vessels and Stripa ground water in an attempt to closely
simulate the repository-like conditions in Stripa.
Several research tools, including solid surface analysis and
solution analysis techniques, were used in combination. These
provided a direct evaluation of nuclear waste glass leaching under
various test conditions.
Major Conclusions
1. A significant compositional effect on glass leaching was
observed under burial conditions. The leach rate expressed by the
annual boron depletion depth was inversely correlated with (SÍO2 +
A^O^/O^O +02O2) wt ratio in the glasses; F^O represents the alkali
oxides.
2. Accelerated attack during the first year in the presence of
bentonite appears to be a transient effect. The presence of
stainless steel, Cu and Ti does not have much effect on glass
leaching.
3. The leach rates of buried samples based on boron depletion
at 90°C ranged from 0.3~3 pm/year for the glass/glass interfaces
investigated. These values are at least two orders of magnitude
lower than those for glasses leached using MCC-1 static leaching
procedures and deionized water.

38
4. Comparison of the laboratory simulation results with field
test results indicates that glass leaching mechanisms were similar
under both test conditions.
5. A model, based on glass structure and thermodynamic
considerations, was proposed to better describe alkali borosilicate
glass leaching than the recent model proposed by Grambow.
6. The results show that Stripa burials combined with
laboratory simulations are unique experimental designs which have
provided useful information regarding nuclear waste glass leaching.
This work has served as a model on which design and development of
the most recent burial test programs are based.

CHAPTER IV
MATERIALS AND METHODS
Glass Compositions and Characterization
Burial Samples
Six alkali borosilicate simulated nuclear waste glass
compositions were used in the burial experiments. They included
three American SRL glasses and three Swedish ABS glasses. Their
compositions are listed in Table 4-1 .
Frit 131 and frit 165, which were designed to contain the
Savannah River Plant (SRP) nuclear waste, were used to prepare SRL
glass samples. Two glasses containing 29.8 wtí and 35 wtí TDS*
waste, respectively, were prepared from frit 131. Another SRL glass
was prepared from frit 165 and contains 29.8 wt? TDS waste. ABS 39
and 41, developed and produced by Dr. T. Lakatos, Swedish Glass
Research Institute, Vaxjo, Sweden, contain 9Í by weight simulated
fission product oxides. These glasses are similar to the COGEMA
glass selected for vitrification of commercial HLW in LaHague,
France, operations [23]. ABS 118 contains 11.25 wtí simulated
fission product oxides and has a composition very close to that of
the future COGEMA glass.
The glass frits were premelted from chemicals using standard
procedures. The simulated SRP waste was mixed with the frit before
* See the notes in Tables 4-1.
39

40
Table 4-1.
Nominal Waste Glass
Burial.
Compositions
(wt?)
Used in the
Stripa
SRL 131 +
SRL 165 +
SRL 131 +
Component
29.8% TDS+
29.8? TDS+
35? TDS+
ABS 39
ABS 41
ABS 118
From glass
frit
Na20
12.4
9.1
11.5
12.9
9.4
9.9
L i ^0
4.0
4.9
3.7
—
3.0
2.0
ZnO
—
—
—
—
3.0
2.5
MgO
1.4
0.7
1.3
—
—
—
AlpOq
—
—
—
3.1
2.5
4.9
b2o^
10.3
7.0
9.6
19.1
15.9
14.0
Fe2Ü2
—
—
—
5.7
3.6
2.9
L
0.4
—
0.3
—
—
—
Si02
40.6
47.7
37.6
48.5
52.0
45.5
Ti02
0.7
—
0.7
—
—
—
Zr02
0.4
0.7
0.3
—
—
1 .0
uo2
—
—
—
1 .7
1.6
0.9
P2°5
—
—
—
—
—
0.3
Cr2Ü2
—
—
—
—
—
0.5
NiO
—
—
—
—
—
0.4
CaO
—
--
—
--
4.0
From simulated waste
Fe^^
13.4
13.4
15.8
—
—
—
Mn02
3.9
3.9
4.5
0.78
0.78
0.97
Zeolite** 2.9
2.9
3.4
—
—
—
A^O^
2.7
2.7
3.2
—
—
—
NiO
1 .6
1 .6
1.9
0.37
0.37
0.47
Si02
1 .2
1 .2
1 .4
—
—
—
CaO
1 .0
1 .0
1.2
—
—
—
Na20
0.9
0.9
1 .0
— •
—
—
Coal
0.7
0.7
0.8
—
—
—

Table 4-1.--continued.
Component
SRL 131 +
29.8% TDS+
SRL 165 +
29.8% TDS+
SRL 131 +
35? TDS+
ABS 39
ABS 41
ABS 118
Nc^SOjj
0.2
0.2
0.2
—
—
—
Cs2C03
0.1
0.1
0.2
—
—
—
SrCO^
0.1
0.1
0.2
—
—
—
U3°8
1 .1
1 .1
1.3
—
—
—
CS2O
—
—
—
0.89
0.89
1.11
SrO
—
—
—
0.26
0.26
0.33
BaO
—
—
—
0.46
0.46
0.58
Y2°3
—
—
—
0.15
0.15
0.19
Zr02
—
—
—
1.29
1.29
1.62
Mo03
—
—
—
1 .65
1 .65
2.06
Ag20
—
—
—
0.01
0.01
0.01
SnO
—
—
—
0.02
0.02
0.02
Sb ^0^
—
—
—
0.004
0.004
0.005
L
—
—
—
0.72
0.72
0.90
Nd203
—
—
—
1.22
1 .22
1 .53
Pr203
—
—
—
0.38
0.38
0.48
—
—
—
0.76
0.76
0.95
CdO
—
0.03
0.03
0.03
Total
100.0
99.9
100.0
100.0
100.0
100.1
+ TDS waste as received from SRL contained Fe203, MnC>2, zeolite,
A120o, NiO, Si02, CaO, Na20, Coal and Na2S0^. This waste was also
doped with Cs, Sr and U.
** Zeolite contains (in wt?) 67.2 Si02. 19.3 A1203, 6.3 Na2Ü, 3.4
Fe202, 2.8 CaO and 1.0 MgO.

42
vitrification. The mixture was fused at 1150-1200°C for 2-6 hours
and annealed at 500-525°C for 1 hour.
Two sample configurations were used: (1) minicans and (2)
pineapple slices. The minicans were made by casting the molten glass
in stainless steel rings 3 mm in diameter by 35 mm long. After
annealing, a hole 200 mm in diameter was drilled through the center
of each minican. Both surfaces of the minicans were polished to a
6-pm finish with diamond paste. Pineapple slices were prepared by
casting cylinders 51 mm in diameter by 80 mm long in molds containing
center carbon posts. Sections 5 mm thick were sliced from the
annealed cylinders and the center posts were removed. One side of
each pineapple slice was polished to a 600-grit (-17 pm) surface
finish while another side was kept in as-cut condition for easy
identification of the glass interfaces after burial. Figure 4-1
shows the pineapple slices of glass, granite,* stainless steel, Ti,
Pb and compacted bentonite** before burial.
Before burial, each sample was subjected to two types of surface
analyses: (1) optical microscopy and (2) Fourier transform infrared
reflection spectroscopy (FT-IRRS). Four to six spots on the polished
surface of each pineapple slice and two spots each on both sides of
the minican were analyzed using these two techniques of surface
* The granite was obtained from Stripa, Sweden.
** The bentonite was obtained from Wyoming. The compacted bentonite
was made by means of isostatic compaction under 100 MPa of
pressure. This is a so-called sodium bentonite whose main
constituent (90 wtí) is montmorillonite.

-C=-
uo
Fig. 4-1. Pineapple slices of glass, granite, stainless steel, Ti, Pb and compacted bentonite before
burial.

44
analyses. In addition, each sample was weighed before burial. Table
4-2 is the sample matrix of the burial experiments.
Glass Quality
Fourier transform infrared reflection spectroscopy (FT-IRRS) was
used as a nondestructive analytical tool for characterization of
glass surfaces prior to the burial. One objective of this
statistical analysis was to determine the relationship between the
FT-IRRS spectra and glass composition used in the Stripa burial. A
second objective was to check if there were any appreciable
variations in composition and/or surface finish conditions among
samples of the same glass formulation. SRL glass samples were used
in this study. The FT-IRRS spectra were obtained on 4 to 6 spots
along the diameter of each glass sample.
Table 4-3 lists the statistical variations of the FT-IRRS
analysis for the SRL glasses. Figure 4-2 shows the representative
spectra of SRL glass pineapple slices before burial. It is observed
that both the wavenumber and the intensity (integrated area under the
curve) for the broad peak containing Si-0 stretching vibrations
increase with increasing SÍO2 content in the glass composition, i.e.,
in the order of SRL 131 + 35% TDS, SRL 131 + 29.8? TDS, SRL 165 +
29.8% TDS. Range of variation in peak position for the same
composition was 0.4-0.8?. The standard deviation of peak position
for the SRL 131 + 29.8% TDS glass slices was the largest (0.8?) due
to glass heterogeneities contained in a few samples of this
composition. On the other hand, the peak intensity and integrated
area under the spectra vary more than peak position because peak

45
Table 4-2. Sample Matrix of the Stripa Burial Tests.
Time
(month)
SRL 131 +
29.8? TDS
SRL 165 +
29.8? TDS
SRL 131 +
35? TDS
ABS 39
ABS 41
ABS 113
Minicans*
It
1
90°C
90°C
—
—
—
--
3
90°C
90°C
—
—
—
—
12
90°C
90°C
—
—
—
--
24
8-10°,90°C
8-10°,90°C
—
—
—
—
Pineapple Slices**
1
90°C
90 °C
90°C
90°C
90°C
—
2
—
—
—
—
—
90°C
3
90°C
90°C
90°C
90°C
90°C
—
4
—
—
—
—
—
90°C
6
90°C
—
—
—
—
—
7
—
—
—
—
—
90°C
12
8-10°,90°C
90°C
90°C
90°C
90°C
90°C
24
8-10°,90°C
8-10°,90°C
8-10°,90°(
—
—
31
--
--
90°C
90°C
--
* Including glass/glass and glass/bentonite interfaces.
** In the case of SRL glasses, glass/glass, glass/bentonite,
glass/granite, glass/Ti and glass/stainless steel were included
with extra two interfaces, glass/Cu and glass/Pb for 1-month
burial; all ABS glasses included glass/glass, glass/bentonite,
glass/granite, glass/Ti, glass/Cu and glass/Pb interfaces.

46
Table 4-3. Variations in Spectral Characteristics of SRL Waste
Glasses.
Glass
Peak
Location
(cm 1 )
Peak*
Intensity (%)
Integrated Area
(Relative Value)
Minicans
SRI 131 +
29.8% TDS
989
±
ij**
21.26 ± 1.67
5.22
SRL 165 +
29.8% TDS
996
+
6
23.12 ± 1.53
5.58
Pineapple
Slices
SRL 131 +
29.8% TDS
980
±
8
21 .71 ± 2.42
5.16
SRL 165 +
29.8% TDS
990
±
6
22.82 ± 2.34
5.44
SRL 131 +
35% TDS
974
±
6
20.16 ± 2.63
4.67
* The compound peak containing Si-O-Si stretching vibrations at
800-1200 cm-1 was used in the statistical analysis.
** Mean and standard deviation.

REFLECTANCE (%)
30.0
WAVENUMBERS
Fig. 4-2. Representative FT-IRRS spectra of SRL glass pineapple slices before burial.

intensity is sensitive to variations in the surface roughness due to
polishing. All these results show that the glass samples except
those containing crystallites are homogeneous compositionally but the
surface polishing conditions have relatively wide variations.
Laboratory Samples
In laboratory simulation tests, most of the glass samples were
made from two similar glass formulations, SRL 165 + 29.8? TDS and
Black Frit 165-Mobay (see Table 4-4). The same melting procedures as
for the burial samples were followed in making the laboratory
glass. The glass melt was cast into a graphite mold. The glass bars
were annealed at 500°C for 1 hour, then furnace cooled.
After cutting from glass bars, samples were polished on all
surfaces up to 600 grit with SiC papers. After cleaning, each sample
was subjected to two kinds of surface analyses: optical microscopy
and FT-IRRS. All samples were weighed before corrosion.
Stripa Field Tests
Sample Assemblies, Minicans and Pineapple Slices
The minicans and the pineapple slices, granite slices, compacted
bentonite slices, stainless steel, Ti, Pb and Cu coupons were
0
assembled at the University of Lulea, Sweden, to provide a wide range
of glass/repository materials interfaces. Minicans were designed to
closely simulate a waste package in a disposal hole. Minicans and
compacted bentonite coupons were stacked together to provide
glass/glass and glass/bentonite interfaces (Fig. 4-3). Sleeves of Pb
and Ti or Cu overpacks were placed around the steel wall of the
minican and a bentonite sleeve separated the waste package from the

49
Table 4-4. Nominal Composition
of Black Frit 165—
Mobay Glass.
Component
Wt %
Si02
55.61
Fe203
11.34
Na20
10.44
B2O3
7.23
L i ^0
4.82
A1203
4.22
Mn02
2.11
CaO
1.10
NiO
0.90
MgO
0.70
Zr02
0.90
F
0.14
Cl
nil
Pb
nil
k2o
0.14
Ti02
0.23
BaO
0.07
ZnO
0.05
Total
100.00
Note: Glass was supplied by
Savannah River Laboratory, Aiken,
SC. This is a similar formula¬
tion to SRL 165 + 29.8$ TDS.
However, it contains more Si02
and less Fe203-

50
Fig. 4-3. A minican assembly

51
walls of the borehole. A typical pineapple slice assembly before
burial is shown in Fig. 4-4.
The SRL glasses included seven pineapple slice assemblies with
different sample stacking sequences (Fig. 4-5 and Table 4-2) and five
minican assemblies with the same sample stacking sequence (Table
4-2).
All the Swedish ABS glass assemblies had the same stacking
sequence to provide six different interfaces (see the footnotes in
Table 4-2). Thus, 35 glass/repository materials interfaces were
involved in these Stripa burial tests with six alkali borosilicate
simulated nuclear waste glass compositions.
Stripa Repository
The Stripa abandoned iron mine was chosen as an underground
field laboratory where the major rock formation is a massive, grey to
light red, medium-grained granite. The mine is located in central
Sweden. The massive and compact nature of granite makes it very
impermeable to water. The hard rock formation has great structural
strength and resistance to erosion or other disruptive events.
Hence, nuclear waste glass placed deep in granite is very unlikely to
be disturbed by climatic or geological events, or by accidental human
intrusion [58].
Table 4-5 lists the average major/minor chemical and mineral
constituents of the Stripa granite. There are several fracture
systems. The majority of the fractures are closed and filled mainly
with chlorite but occasionally with calcite. This mine provides an
environment which closely simulates an actual granite repository and

52
Fig. 4-4. A typical pineapple slice assembly.

Fig. 4-
Pre- Burial
12 24
6 iz
10* C
U1
UJ
5. Seven preburial pineapple slice assemblies with different sample stacking sequences for SRL
simulated nuclear waste glasses.

54
Table 4-5. Average Major/Minor Chemical
and Mineral Constituents in
Stripa Granite.
Oxide
Wt %
Si02
74.7
Al^jO^
13.2
Fe20n
1.6
FeO
NR a
MgO
0.20
CaO
0.6
Na20
4.0
k2o
4.6
Ti02
0.05
P2°5
NR
MnO
0.03
BaO
0.02
h2o
NR
OJ
o
o
NR
Mineral
Grey
Red (vol %)
Quartz
33
44
K or Na Feldspar
24
12
Plagioclase
35
39
Biotite
<1
NR
Muscovite
<1
2
Chlorite
<1
3
Adapted from [59].
a NR = not reported.

55
is thus ideal for conducting glass corrosion experiments. The
location within the Stripa mine where the samples were buried is
shown in Fig. 4-6. This is about 345 meters below the surface. The
holes into which the samples were placed were about 3-m deep and
56 mm in diameter, which were filled with ground water from the mine
before the samples were placed to a depth of about 2.5 m (see Fig.
4-7). The arrow indicates holes for 1-month, 90°C specimens. The
ground water composition and pH prior to burial as measured in the
recent study is given in Table 4-6. Table 4-7 lists the ground water
composition and pH found in literature [60],
Burial and Retrieval
All the sample assemblies were buried in the boreholes at the
345-m level below the surface. Heater rods were placed in the center
20-mm holes on the samples designed to be maintained at 90°C (Fig.
4-6) to simulate the thermal period of -300 years. Without the
heating elements, glass samples were tested at ambient mine
temperature (8-10°C), which is expected to be the temperature of a
canister in a real repository after about 300 years. As shown in
Fig. 4-7, a rubber seal was used to prevent water intrusion from the
floor of the mine. Water entering the hole had to permeate through
the granite.
Assemblies were retrieved at specific intervals over a 3_year
period (see Table 4-2). After removal of the burial assemblies from
the boreholes, they were wrapped in plastic until disassembled and
analyzed. Water from the boreholes was analyzed prior to and after
the assemblies were removed.

56
BURIAL SITE AT STRIPA
GRANITE
SHAFT
MINE TUNNEL
íR capped
HOLE
Fig. 4-6. Location within Stripa where SRL samples were buried.
This is about 345 m below the surface. The holes into
which the samples were placed are about 3~m deep and 56 mm
in diameter. They were filled with water from the mine
before the samples were emplaced. The arrow indicates
holes into which the 1-month, 90°C specimens were
placed. Samples were placed in the hole to a depth of
about 2.5 m.

57
SCHEMATIC OF BURIAL ASSEMBLY AT STRIPA
GRAN1U
345 m*fer
Section of S^mplr
(Pr* Burial I
Fig. 4-7. Diagram illustrating the position of the samples in the
Stripa mine during burial. A pineapple slice and minican
are also shown, along with a photograph of the 1-month,
90°C assembly immediately after removal from the borehole.

Table 4-6.
Ground Water Composition and pH Measured in This Study within the 1-month Test Hole at
Stripa. Concentration mg/L.
pH Li B Na Al Si Mn Fe Zn Sr Mo La Nd ^
CO
Before Burial 8.1 0.16 <1 30 0.02 7.7 <0.1 <0.1
0.30 0.16 <0.01

59
Table 4-7.
Ground Water Composition for
the Stripa Granite, Literature
Values.
Anions
mg/1
hco3"
Cl"
so42"
15.4-78.7
52-283
2.7-1.9
f"
NRa
Cations
Ba2+
Ca2+
Fe3 +
NR
10-59
0.02-0.24
Li +
Mg2 +
K+
NR
0.5
0.2-5.4
Si02
Na+
Sr2+
11 .0-12.8
43-125
NR
pH
8.85-9.75
Total Dissolved 200-230 (330-410 m)b
375-510 (below 700 m)b
Adapted from [60].
a NR = not reported.
b Depth of sample below surface.

60
The measured flow rates through the boreholes near those where
the assemblies were located were approximately 1 L/year (0.1 mL/hr)
[61]. The glass surface area to ground water volume ratios (SA/V)
were estimated to be >1 cm 1 and were most likely different from spot
to spot on some of the samples due to different water accessibility
at the glass interfaces. The calculated SA/V ratios were low, about
0.6 cm 1 for the pineapple slices and 0.06 cm 1 for the minicans.
This was based on the volume of water below the rubber seal and the
total surface area of the glass in the hole.
The postburial procedures consisted of careful disassembling,
soaking in deionized water for no more than 5 min to remove excess
bentonite, if present, and two to three 5-min ultrasonic cleanings in
acetone or absolute ethanol. The samples were air-dried and placed
in a desiccator until analyzed.
Disadvantage of the Burial Test Method
The primary disadvantage of the Stripa burial is that it was not
possible to calculate leach rates based on the data of solution
analyses. This is because, when the samples were taken out of the
borehole, the ground water above the rubber seal ran into the lower
part of the borehole where the sample assembly was positioned. In
addition, other contaminants may be present in the ground water. All
these make the leach rate calculation based on the solution analysis
data meaningless. Therefore, surface analyses had to provide the
primary evaluation method for assessing glass performance and for
comparing field- and laboratory-corroded specimens. As will be
mentioned, the Materials Interface Interactions Tests

61
surface/solution analysis (MIIT-SS) effort contains an improvement
over the Stripa burial in that solution analyses will be obtained
[62].
Similar Tests Being Used in MIIT Studies at WIPP
The Materials Interface Interactions Tests (MIIT) is a series of
experiments that will assess the performance of simulated SRL waste
glass along with a variety of additional simulated waste glass
compositions in the presence of various proposed canister, overpack
and backfill components, in the salt geology at the Waste Isolation
Pilot Plant (WIPP) [62]. Design and development of the MIIT tests
were derived from the experience obtained through in-situ testing of
over 100 simulated SRL waste glass samples buried in Stripa granite,
Sweden. The MIIT in-situ testing program represents a "second
generation" of the Stripa tests. The MIIT studies consist of two
parts, MIIT-MI (multiple interactions), and MIIT-SS (surface/solution
analysis). The MIIT-MI effort is similar to the Stripa experiments
and involves glass performance as a function of a variety of proposed
package components, predominantly by surface analyses. The MIIT-SS
effort represents a significant improvement over the Stripa burial
experiments in that solution analysis will also be obtained for
simplified interactions, and time-dependent data will be obtained
from single boreholes. Only pineapple slice assembles will be
utilized. All tests will be conducted at 90°C. Samples will be
removed from the mine at time intervals of 6 months, 1 year, 2 years
and 5 years.

62
Laboratory Tests of Simulated Corrosion
In order to simulate the actual repository conditions, two sets
of laboratory leaching tests were conducted. In one set, a modified
MCC-1 static leach test method was used for SRL 165 + 29.8í TDS glass
with two different glass surface area-to-volume of leachant (SA/V)
ratios, 0.1 and 1.0 cm 1. The leachant was selected from one of the
following: deionized water, Stripa ground water and Stripa ground
water saturated with glass powders of the same composition as the
bulk specimen at 90°C for 14 days. Prior to immersing in the
leachant, each specimen was ultrasonically cleaned in either reagent
grade acetone or absolute ethanol for 3 times, 5 min each. The
samples were suspended inside either a PFA Teflon* (60 ml capacity)
corrosion cell and then placed inside a constant Blue M** convection
oven as shown in Fig. 4-8.
In another set of so-called "rock cup tests," a Stripa granite
cup was placed in each PFA Teflon container to simulate granite
repository conditions. The granite was obtained from boreholes in
Stripa, Sweden, and the granite cups were made by Diversified Machine
Works, Post Falls, Idaho. A diamond drill was used to drill a hole,
3.2 cm in diam. by 3-8 cm deep in each granite cylinder, 4.4 cm in
outside diam. by 4.9 cm high. Monolithic glass samples of Black Frit
165-Mobay (see Table 4-4) were placed in the cup. A certain volume
of ground water was filled both inside and outside the cup. Some of
* 0102-53 MOD PFA Teflon jar, Savillex Corp., Minnetonka, MN.
** Model OV-490A-2, Blue M Co., Blue Island, FL.

ENVIRONMENTAL TESTING SYSTEM
Fig. 4-8. Schematic of experimental configuration of static leach test.

64
the cups contained stainless steel (316 L) wires used for supporting
the glass specimens. The rock cups were soaked in ground water for 2
days, then air dried.
Both static and flow test conditions were used in the rock cup
tests. The SA/V ratio was 1.0 cm 1 in the cells. In the case of
2
static leaching, the surface area of glass sample was about 14 cm .
A stainless steel (316 L) wire, 0.1 cm in diameter by 18 cm in
extended length was contained in the rock cup. All glass samples and
stainless steel wires were cleaned ultrasonically 3 times, for 5 min
before leaching with absolute ethanol. A corrosion cell for the rock
cup static test is similar to that for the flow test shown in Fig.
4-9 but without the fittings in the lid of the Teflon container.
In the rock cup flow test, low flow rates, 0.1-0.3 mL/h, were
used. This single pass continuous flow test method was similar to
the MCC-4S procedures [27]. A stainless steel (316 L) wire, 0.1 cm
in diameter x 36 cm in extended length, was contained in the rock
cup. The glass surface area was about 26 cm . The procedures of
sample preparation and granite cup cleaning were the same as in the
static test. A flow leaching vessel and the experimental set-up are
shown in Figs. 4-9 and 4-10. Only the leachant within the cup was
forced to flow using a Peristaltic cassette pump.* The flow rate of
the ground water was controlled to ±10? of the set value. The ground
water was not preheated in the reservoir. Since the flow rate was
low and the ground water prior to being introduced into the leaching
* Made by Manostat, New York, NY.

65
Fig
4-9
A corrosion cell in the flowing test

BLUE M CONVECTION
Fig. ^-10. Schematic of experimental configuration of continuous flow test.

67
vessel was kept at 90°C in the tubing for some time (longer than 5
min), the temperature within the leaching vessel would not be changed
due to the water flow. The leachant after passage through the leach
vessel was collected weekly.
All the laboratory tests were run at 90°C for up to 6 months.
The sample matrix of the experiments is shown in Table 4-8.
Before leaching, each sample was weighed and examined under an
optical microscope. FT-IRRS was run at 2-3 spots on each sample.
Analytical Techniques
A combination of several analytical techniques was used for
evaluating nuclear waste glass leaching. Each of the methods yields
averaged information which is characteristic of a volume extending
from the surface to a specific depth within the sample, as shown in
Fig. 4-11 and Table 4-9. In addition, SIMS provides depth resolved
concentration profiles from the surface into uncorroded bulk. As
discussed earlier, all the solid surface analysis techniques in Fig.
4-11 have been used for characterizing changes on the Stripa burial
glass surfaces. Solution analysis techniques including inductively-
coupled plasma (ICP), atomic absorption spectrophotometry,
colorimetry, and pH measurement were also used with the laboratory
leached specimens.
Solid State Analyses
Optical microscopy
Each sample was examined under a microscope using the reflection
light mode, both prior to and after leaching. Magnification of 100X

68
Table 4-8. Sample
Matrix of the Laboratory Tests.
Glass Composition*
SRL 1 65 + 29.8% TDS, Black Frit 165-Mobay
Temperature
90°C
SA/V
0.1 and 1.0 cm 1
Leaching Time
1 , 3 and 6 months
Leaching Condition
Static and flow (0.1 and 0.3 mL/h), with and
without granite cup
Leachant
deionized water
Stripa ground water
Stripa ground water saturated with glass
powders for 14 days at 90°C
*
Samples were run in duplicates.

69
RBS
¡S|GHT SEM-EDS
MICROSCOPE FT-IRRS SiMS/lON MILLING
5-200 A
0.5>¡m
I.Sjjm
BULK GLASS
SOLUTION ANALYSIS
ALTERED
LAYER
V
Fig. 4-11. Sampling depths with various techniques used in this
study (adapted from [63,64]).

70
Table 4-9. Characteristics of Analytical Techniques.
Sampling Spatial Detection
depth resolution Information limits(?)
Secondary ion mass
spectroscopy (SIMS)
5-20A
(profiling
to =10pm
100A-=1pm
composition
structure
<1 0
Fourier transform
infrared reflection
spectroscopy (FT-IRRS)
=0.5 pm
3~5mm
composition
structure
morphology
3
Scanning electron
microscopy-energy
dispersive spectroscopy
(SEM-EDS)
1.5pm
1 . 5pm
morphology
composition
5
Rutherford back
scattering (RBS)
100 A~=1pm
1 mm
composition
>10~3

71
was used in all cases, which permits examination of the general
surface characteristics. For preleached glasses, heterogeneities,
such as crystallites, can be observed (Fig. 4-12). These
heterogeneities usually result when glass homogenization is not
complete or the wastes have not been dissolved by the glass matrix.
The glass surface finish conditions also can be checked (Fig.
4-13). In this study, glasses were polished to 600 grit or 6-pm
surface finish. Examination with an optical microscope served as a
quality control for the sample conditions. For the leached glass
surfaces, both surface roughening and surface precipitates can be
evaluated using this simple, rapid, and inexpensive technique.
Fourier transform infrared reflection spectroscopy (FT-IRRS)
Fourier transform infrared reflection spectroscopy (FT-IRRS) has
recently been developed as a semi-quantitative tool for
characterizing the surface structure and composition of glasses both
prior to and after exposure [65,66]. The important advantages of
this technique include (1) it does not require vacuum and energetic
electron or ion bombardment; thus it does not alter the surface of
the glass as may Auger electron spectroscopy (AES), electron
spectroscopy of chemical analysis (ESCA) and secondary ion mass
spectrometry (SIMS); (2) it is applicable to in-situ glass surfaces
of nearly any configuration and can be used for analysis of large or
small areas, if desired, is relatively inexpensive and requires only
standard infrared spectrometers; and (3) the FT-IRRS method can be
used as an automated analytical tool and can also be coupled with
solution analysis, making it especially suitable for characterization

72
Fig. 4-12.
Light micrograph of SRL 131 + 29.8? TDS glass with
crystallites (100X).

73
Fig. 4-13.
Light micrograph of a typical glass surface
polishing to 600 grit surface finish (100X)
after

74
of surface/environment interactions [67-69]. In a spectrum of the
binary soda-silica glass surface, the region where the Si-O-Si
stretching peak (S) and silicon-oxygen-alkali (NS) stretching peak
overlap occurs is called the coupled region. Exposure of the glass
to a chemical environment alters the relative concentration of both
silica and alkali ions due to preferential leaching of the alkali
ions. This produces the decoupling of the S and NS peaks in the
infrared reflection spectra.
Extensive surface reactions can lead to roughening of the glass
surface due to formation of either pits or surface deposits.
However, the wavenumber location of the S and NS peaks is not changed
significantly by the surface roughening. Therefore, it is possible
to use the shift of the wavenumber location of the FT-IRRS peaks to
measure the change in composition of the glass surface, independent
of roughening or surface deposition. The extent of surface
roughening can be assessed by the decrease in intensity of the
FT-IRRS peak when wavenumber location remains unchanged.
It should be noted that the FT-IRRS technique is useful for
determining changes in reaction layers of -0.5 pm thick or greater.
Because of the 0.5-pm sampling depth, the information collected by
this technique within the sampling depth is averaged and accurate
analysis of very thin (<1000 A) surface corrosion films is not
possible. For very thick reaction layers, FT-IRRS provides an
analysis of only the outer -0.5 pm of the layer using near normal
specular reflectance. This nondestructive testing technique is
valuable for quick and efficient routine controls while in other

75
cases more detailed and usually more expensive analyses such as SIMS
are necessary. Figure 4-14 shows the FT-IRRS analyses of SRL 165 +
29.8? TDS glass/glass interface prior to and after 2-year burial in
Stripa at 90°C. The decoupling of the S and NS peaks in the region
of 800-1150 cm-1 and loss of peak intensity as shown in the
postburial spectrum are a result of leaching.
Scanning electron microscopy/energy dispersive
spectroscopy (SEM-EDS)
The major advantages of SEM over other techniques such as
optical microscopy are that much higher magnification and a greater
depth of field are possible. The specimens were usually vacuum
coated with 100 A of C or Au-Pd. The information obtained using this
technique is mainly qualitative, although EDS in favorable cases may
yield the average composition of the outer most few microns. Figure
4-15 shows a typical SEM* micrograph of a glass surface after
polishing to 600 grit and prior to leaching. Figure 4-16 shows the
EDS data of SRL 131 + 2.98? TDS glass prior to burial.
Secondary ion mass spectroscopy (SIMS)
Secondary ion mass spectroscopy (SIMS) has an information depth
of the order of one atomic layer combined with ionic milling
(sputtering), which together with high detection sensitivity for most
elements offers a unique potential in profiling. During the last
several years, advances have been made by scientists at the Chalmers
* Scanning electron microscope, model JSM-35CF, JE0L Ltd., Tokyo,
Japan.
I

30
20 -
Lü
O
Z
?
ti
_l
Li_
Lü
tr
10
/ \
/ \
S /
fs
1/ \
> \
t \
Pre-burial
\
NS
\ ^ Post -burial
90°c
\\
\\
\ \
\ \
\ \
\ .
I
1200
1000 800
WAVENUMBER (cm"')
600
Fig. 4-14. FT-IRRS analysis of SRL 165 + 29.8% TDS glass/glass interface prior to and after 2-year
burial in Stripa. The postburial spectrum shows the decoupling of the S and NS peaks in
the region of 800-1150 cm-1 and loss of peak intensity as a result of leaching.

Fig. 4-15. SEM micrograph of a typical glass surface after polishing
to 600 grit prior to leaching.

RELATIVE INTENSITY
KEV
Fig. 4-16. EDS analysis of an uncorroded SRL 131 + 29.8% TDS glass surface.

79
University of Technology, Sweden, to develop SIMS as a sensitive and
routine tool in the study of glass corrosion [70]. The glass samples
were coated with a 100 A Au film to reduce surface charging. The
Cameca 3-F ion probe accelerated and focused a beam of 0 ions
towards the glass sample, successively eroding the surface by
sputtering, while cyclically counting the yields of sputtered
secondary ions of different species which can be detected and
quantified with a mass spectrometer. The raw data, processed by an
on-line computer, consisted of these ionic yields vs the
corresponding sputtering time. With the aid of known relative
elemental sensitivity factors (RSF), the ionic yields were converted
to the percent atom concentrations of all the measured elements and
their sum of cations was set equal to 100 percent. Although H is
measured, it is not included in the conversion calculation, because
the H content of the preleached glass is unknown. The determination
of relative erosion speeds at different depths of the sputtered layer
permitted the conversion of sputtering time to depth. As an example,
the element concentration profiles of ABS 118 glass/glass interface
after 12-month, 90°C burial are shown in Fig. 4-17.
In the most recent version, the profiles were corrected to
consider the elemental release and absorption during corrosion.
These profiles are different from the atomic concentrations (in
percent) shown in Fig. 4-17. The new profiles indicate the actual
gram«atoms of each element after leaching of 100 gram«atoms of
original glass, and so may be directly used in calculations of
elemental losses. Due to the cation (except H) release and

ATOMIC CONCENTRATION (%)
80
Fig. 4-17. SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 90°C burial in Stripa. The atomic
concentrations of all cations (except H) are summed up to
100?.

81
adsorption, these actual gram^atoms of various cations may not be
summed up to 100 after leaching. These profiles were calculated
using the least leachable elements (in most cases Al, sometimes Fe,
Mn or Zr) as a standard and assuming that their actual gram*atoms
remain unchanged at any time at the surface.
Taking Al as a standard element, the following equation can be
wrriten based on 100 gram*atoms of unleached glass at a specified
depth in the glass surface,
GA.. _ - GA + GA " GA... , (4
Al,after Al,before Al.abs Al,leached
where GAftl after = gram*atoms of Al after leaching;
GAA1, before = 8ram*atoms of Al before leaching;
GAAi at3S = gram«atoms of Al absorbed from solution;
GAA1, leached = Sram’atoms of Al leached.
Since the concentrations of Al in the ground water were low both
before and after leaching (see Table 5-1), neglecting the last two
terms on the right side of equation (4-1) will not introduce
appreciable error. Thus, we have
GA.. „. — GA.. , .
Al,after Al,before
(4-2)
where GAftl after and GAA1 before are same as in equation (4-1).
Also, we can write

82
GAA1,after at'^Al,after ^ ^i,after
i
(4-3)
where at.?A1 after is concentration of A1 (in at.?) at a certain
depth after leaching as shown in Fig. 4-17 and E GA. is a
. i,at U6P
i
summation of the gram-atoms of element i after leaching. Combining
equations (4-2) and (4-3) gives
GA ^ — at.? • E GA.
Al,before Al,after . l,after
i
(4-4)
Also
GA. = at.?. • E GA.
i.after i,after . i,after
i
(4-5)
where at.?^ after is concentration of element i (in at.?) at a
certain depth after leaching, as shown in Fig. 4-17. From equations
(4-4) and (4-5), we have
at.?
GA.
i .after
GA
Al .before
i.after
at. ?
(4-6)
Al.after
Using equation (4-6), the actual gram-atoms of each element left
based on 100 gram-atoms of unleached glass at a certain depth can be
calculated. The results as obtained from the on-line computer of the
SIMS instrument are given in Fig. 4-18 for the same glass specimen
shown in Fig. 4-17.

GramatDms Remaining Based on
IOO Gramatoms of Unleached Glass
83
Fig. 4-18. SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 90°C burial in Stripa. Data are presented
as granratoms of various cations remaining in the leach
layer at certain depth based on 100 gram*atoms of
unleached glass.

84
Density calculation. The density of the leached surface as a
function of depth was estimated using SIMS depth profiles of
concentration (such as Fig. 4-17), based on conservation of matter
and assuming that the volume concentration of the least-soluble
species, such as Fe, Ni, Al, Zr, and Mn remain unchanged throughout
the leached layer. Thus
wt. ?.
d
wt. ?
i .bulk
dbulk
(4-7)
where i is Fe^^, NiO, Al-jO^, Zr02 or Mn02; wt?if the wt? of the ith
oxide; d, the density of the leached layer at a specific depth, and
dfauik is the glass density. From equation (4-7),
wt?.
Xd =
i ,bulk
bulk
wt?.
(4-8)
where Id is the density index at a certain depth.
Rutherford back scattering surface analysis (RBS)
The Rutherford back scattering (RBS) method is a near surface
analytical technique with a depth resolution of approximately 20 nm
and a penetration range of up to 4000 nm [71,72]. The method is
generally most effective for heavy elements in a lighter matrix.
Consequently, RBS analysis may be especially useful for determining
changes in uranium surface concentration during burial and in
evaluating glass/metal overpack interactions such as glass/Pb
interfaces. Furthermore, RBS is practically nondestructive, rapid

85
(15 min/spectrum) and less expensive when compared with techniques
such as SIMS. The RBS technique is not ideal as a preliminary mode
of analysis on complex specimens with unknown composition because
mass discrimination decreases as the atomic number increases.
However, it is valuable for rapid comparison of complex specimens for
which basic elemental data are well established.
The RBS data were obtained using the van de Graaf generator and
RBS collection system of the University of Florida's Major Analytical
Instrumentation Center. The RBS analysis was performed using a 2 MeV
alpha particle beam. The beam current was -100 nA. The beam area
was -1 mm . The scattering geometry was that of normal incidence
with a scattering angle of 170°. The silicon detector had a surface
p
area of -20 mm and was positioned at -120 mm from the specimen.
Only selective glass samples were run by the RBS analysis.
Solution Analyses
The solutions were analyzed in a variety of ways to obtain
elemental concentrations and pH. These include colorimetry*,
inductively coupled plasma atomic emission spectrometry (ICP)** and
pH determination.*** All the solution analyses and pH measurements
were done at room temperature.
* Hach DR/2 Spectrophotometer, Hach Company, Loveland, CO.
** Inductively coupled plasma atomic emission spectrometer, model
Plasma 200, Instrumentation Laboratory, Walthan, MA.
*** Digital ionalyzer, model 801A and Ross combination electrode,
model No. 810200, Orion Research, Inc., Cambridge, MA.

86
For most of the experiments, ICP was used for analyzing all the
elements. In some cases, Si concentration was determined by using
colorimetry. In an ICP, the sample solution is introduced into a
spray chamber with the help of a nebulizer. The nebulizer converts
the liquid sample into a fine aerosol. Argon gas flows through a
quality torch which carries the smaller aerosol droplets of sample
into the conical channel of plasma. The high temperature of the
plasma produced by a radio frequency generator, desolvates the
droplets and dissociates them into individual atoms and ions. They
are excited to emit light at various wavelengths characteristic of
the element, which are analyzed by a spectrometer. The spectral
intensities are directly proportional to the elemental
concentrations. The photomultiplier converts the light energy into
electrical signals which are then digitized and processed by the
instrument's computer.

CHAPTER V
TEST RESULTS
Field Test Results
General Observation
When the assemblies were removed from the boreholes, there was a
thin (<5 mm) coating of bentonite over their outer surfaces, as shown
in Fig. 5-1. Apparently, the compacted bentonite swelled during
burial and was extruded between the sides of the granite boreholes
and the surfaces of the assemblies. In some instances, bentonite
intrusion was observed between the minicans and the pineapple
slices. For up to 1-year burial, accelerated attack was usually
found on the glass surfaces exposed to compacted bentonite, or where
bentonite had intruded between two samples.
Cracks were observed on many of the pineapple slices during
disassembly. Two possible causes for the cracks were (1) pressure
due to the swelling of the bentonite during burial, which produced a
bending moment on the glass samples, and (2) large forces required to
extract the assemblies from the boreholes (also due to the swelling
of the bentonite), which also produced a bending moment on the
glasses.
The pH changed by less than 1 unit in the boreholes from which
the 1-month, 90°C minican assemblies of SRL glasses and the 31-month,
90°C pineapple slice assemblies of ABS glasses were removed.
87

88
Fig. 5-
Pme.App !â– 
SI
ce.
A "S5e.m b ly
Post Ear\ p.\ - 1 Month- ^CTC
1. A typical assembly after burial in Stripa mine. Bentonite
coating can be observed on the outer surface due to
bentonite swelling.

89
Measured flow rates through similar holes located elsewhere in the
mine were approximately 1 L/yr [61].
Several types of surface areas were observed on the postburial
glass surface. As in the case of glass/glass interface shown in Fig.
5~2, these include clear, cloudy and bentonite-intruded areas.
Variations in corrosion did occur from spot to spot over the same
samples and this is attributed to local water accessibility,
bentonite intrusion and extent of interfacial contact between the
various samples. The actual SA/V ratio for the glass surfaces
(including both polished and unpolished surfaces) are expected to be
higher than 0.03-0.6 cm-1, which was calculated based on the total
area of glass surface in the assembly and the volume of ground water
in the corresponding borehole.
Table 5-1 shows the ground water compositions collected from the
boreholes where SRL glass pineapple slice assemblies were removed.
The concentration of Na increased from 40 to 255 mg/L and that of Si
increased from 2.3 to 35.3 mg/L. The increased Na and Si
concentration in the ground water could be due to leaching from the
glass as well as release from the bentonite with the surrounding
granite.
Results with ABS Glasses
Figure 5-3 compares the FT-IRRS spectra of the glass/glass,
glass/granite and glass/bentonite interfaces for ABS 39 and 41 before
and after 31-month, 90°C Stripa burial. This figures shows that the
postburial spectra of ABS 41 glass/glass and ABS 41 glass/granite
interfaces changed less than those of ABS 39. However, spectra of

90
h2o
H20 accessibility and/or
Bentonite intrusion
Cloudy area slightly
corroded
_ Clear area
uncorroded
- Bentonite intrusion
severely corroded
Fig. 5-2. Schematic of glass/glass interface illustrating several
types of surface areas resulting from water and/or
bentonite intrusion.

91
Table 5-1. Composition of Ground Water Collected from the Boreholes
where SRL Glass Pineapple Slice Assemblies Had Been
Buried. Potassium and Ca are noted but not measured.
Concentration (mg/1)
Postburial (months)
Element
Preburial
1
3
24
Na
30
70
190
1 47
Li
0.16
0.52
0.96
0.17
Si
7.7
10
43
14
A1
0.02
0.04
0.7
0.2
B
<1
0.9
5.48
1 .26
Mo
0.16
0.35
0.29
0.25
Cr
-
-
<0.01
0.06
Fe
<0.1
<0.2
0.5
0.8
Ni
-
-
<0.01
0.11
Mg
-
-
2.13
0.77
Cu
-
-
<0.01
0.01
Zn
<0.1
<0.01
0.13
0.13
Mn
<0.1
<0.2
<0.01
0.03
La
<0.01
<0.01
<0.01
<0.01
Nd
<0.01
<0.01
<0.1
<0.1
Sr
0.30
0.20
0.19
0.38
Ba
-
-
0.20
0.07

FT IRRS spectra of glass ABS 39 (a) and ABS 41 (b) before and after 31-month, 90°C Stripa
burial. y
Fig. 5-3.

93
the glass/bentonite interface did not show much difference between
these two compositions.
As can be seen in Fig. 5~3 (a), all three postburial spectra of
ABS 39 lost considerable intensity. Based on previous FT-IRRS
investigations of glass-water interactions [65,69] and the SIMS
analysis shown in Fig. 5-4, this loss is due to extensive sodium and
boron depletion, surface roughening and probably network
dissolution. For glass/glass and glass/granite interfaces, the
relative silica concentration at the glass leached layer increased,
as indicated by the shifting of the peak associated mainly with
Si-O-Si stretching vibrations from 1010 cm 1 to 1060-1070 cm-1. For
the ABS 39 glass/bentonite interface, no obvious shifting of this
peak to higher wavenumbers was found.
The ABS 41 glass/glass and glass/granite interfaces behaved
quite differently from those of ABS 39. For example, the FT-IRRS
spectrum of the glass/glass interface almost retains the shape of the
preburial spectrum, indicating very little surface attack. This is
confirmed by the SIMS depth profiles. As shown in Fig. 5~4(b), the
leach depth of the ABS 41 glass/glass interface is less than 0.4 pm,
which is below the sampling depth of FT-IRRS (-0.5 pm). In the case
of the ABS 41 glass/granite interface, extensive depletion of alkali
was observed, as shown by the loss in spectral intensity at the lower
wavenumbers in Fig. 5-3 (b). However, the intensity of the major
peak at 1080 cm-1 of this spectrum is as high as that of the
preburial one, indicating very little network dissolution and surface

GRAM-ATOMS REMAINING BASED ON
IOO GRAM-ATOMS OF UNLEACHED GLASS
94
(a)
Fig. 5-4. SIMS depth profiles for (a) ABS 39 (Al-corrected) and (b)
ABS 41 (Si-corrected) after 31-month, 90°C Stripa burial.

Gram-Atoms Remaining Based on
IOO Gram-Atoms of Unleached Glass
95
(b)
Fig. 5-4.—continued.

96
roughening. Also, the peak originally at 1020 cm 1 in this spectrum
has been shifted to 1080 cm 1 due to an increase in the relative
silica concentration at the alkali- and boron-depleted glass
surface. The FT-IRRS analysis also shows that, among the three
interfaces, ABS 41 glass/bentonite is the worst case with a spectrum
quite similar to that of ABS 39, as discussed earlier.
The optical micrographs of the glasses with three interfaces
after 31"month, 90°C Stripa burial are shown in Fig. 5-5. As can be
seen from these pictures, extensive glass leaching made the ABS 39
glass surfaces very rough. In contrast, the ABS 41 glass surfaces,
except those exposed to bentonite, still appear quite smooth and
clear, showing very little surface attack.
It should be noted that these postburial glass surfaces are
generally covered with corrosion products. For ABS 41 glass/glass
and glass/granite interfaces, the surface film is very thin and in
some areas the glass appears unaltered. For all three ABS 39
interfaces and the ABS 41 glass/bentonite interface, the surface
films formed on the glasses are heterogeneous. Sometimes there seems
to be a layer of either bentonite or superposed flakes of peeled-off
reacted glass on the glass surface, as shown in the ABS 41/bentonite
interface (Fig. 5-5 (f)).
The SIMS depth profiles of the two glasses reveal both the depth
of leaching and the concentration of chemical species in the surface
layers as in Fig. 5-4. Here, to begin with, only the results for
boron will be presented, to indicate mainly the effective depths of
leaching.

97
Fig. 5-5. Light micrographs (100X) of glass 39 (a) glass/glass, (b)
glass/granite and (c) glass/bentonite interfaces and glass
ABS 41 (d) glass/glass, (e) glass/granite and (f)
glass/bentonite interfaces after 31-month, 90°C Stripa
burial.

98
Figure 5-6 compares the boron profiles for the two glass
compositions with three postburial interfaces. The reasons for
selecting boron for measuring the leach depth in this dissertation
are that (1) boron is one of the major glass former ions and the
depth of its depletion provides a good measure of the alteration of
glass network; and (2) in most cases, no phase was known to
precipiate boron from solution. Therefore, its mass loss should
directly indicate the amount of the glass network that had altered.
The ABS 39 glass generally has thicker leached layers than ABS 41, as
indicated by the depth of the boron extraction. For example, after
31 months, the leach depths of ABS 39 glass/glass and glass/granite
interfaces are approximately 18 and 32 pm, respectively, whereas
those of ABS 41 are less than 1 pm. However, the glass/bentonite
interface did not show much difference in leach depth between ABS 39
and 41. In fact, ABS 41 glass leached to a depth of 20 pm as
compared with 15 pm for ABS 39 glass when they were exposed to
bentonite during the 90°C, 31-month burial. Thus, although the
surface of ABS 39 exhibited a more severely attacked aspect, the
long-term leaching depth of ABS 41 surprisingly exceeded that of ABS
39. In the case of glass ABS 39, however, the largest leach depth
was found with the glass/granite interface, not with the
glass/bentonite interface after 31-month burial at 90°C.
Figure 5~7 compares the FT-IRRS spectra of ABS 118 before and
after exposure to either glass, granite or compacted bentonite during
2- and 12-month 90°C Stripa burial. As discussed earlier, selective

99
100
<
cu
f—
2:
Lü
O
2
O
o
o
o
1-
<
10
0.1
o
(b)
ABS 41
STRIPA BURIAL
31 Mo., 90° C
/r
s'
^^-Glass / Glass
r'
/
/
-—Glass/ Granite
/
/
)
/
/
/
/
y
_ _ y
""" Glass/bentonite
""
i I 1
1 1 1
10 20
DEPTH (um)
30
Fig. 5-6. SIMS depth profiles of boron for glass ABS 39 (a) and
glass ABS 41 (b) after 31-month, 90°C Stripa burial.

WAVENUMBERS (cm'') WAVENUMBERS (cm'')
Fig. 5-7. FT-IRRS spectra of glass/glass, glass/granite and glass/bentonite interfaces for nuclear
waste glass ABS 118 buried in Stripa at 90°C for (a) 2 months and (b) 12 months. Also
shown is the spectrum of a preburial glass surface.
100

101
leaching of soluble species such as boron and alkalis results in an
increase in the relative Si concentration at the leached surface and
splitting of the broad peak at the 800-1100 cm 1 region. The S peak
then shifts to higher wavenumbers due to enrichment of Si02- In Fig.
5-7 (a), all the 2-month postburial spectra show the S peak at -1060
cm 1 for the glass/glass and glass/granite interfaces and at the
-1050 cm 1 region for the glass/bentonite interface, as compared to
the broad peak at -1010 cm 1 in a preburial spectrum. The high
intensity of the S peak, 27$ for the glass/glass and 25$ for
glass/granite, indicate slightly roughened surfaces as shown in Fig.
5~7 (a). The glass/bentonite interface, however, shows a rougher
morphology with low S peak intensity (15$) after 2-month burial.
This is consistent with the optical micrographs shown in Fig. 5-8
(a-c) for the three glass interfaces.
As regards the NS peak, the spectrum for the 2-month glass/glass
interface (Fig. 5~7 (a)) shows that there still remains a
considerable amount of alkalis within the 0.5 pm sampling depth of
FT-IRRS. Actually, Si enrichment (as compared to other elements)
within the altered glass surface also resulted from leaching of
species other than alkalis, such as B. The SIMS data shown in Fig.
5-9 (a) confirm this concept by showing that B depletion within the
0.5 pm outer surface was nearly complete. For the glass/bentonite
interface, the FT-IRRS analysis shows that within 0.5 pm depth, most
alkalis have been leached out during 2-month 90°C Stripa burial.
After 12-month 90°C Stripa burial, the FT-IRRS data (Fig. 5-7
(b)) indicate that the roughness of the glass/glass interface

102
Fig. 5-8. Light micrographs (100X) of glass ABS 118 after 2-month
burial, (a) glass/glass, (b) glass granite, and (c)
glass/bentonite interfaces, and after 12-month burial,
(d) glass/glass, (e) glass/granite, and (f)
glass/bentonite interfaces at 90°C in Stripa.

Gramatoms Remaining Based on IOO
Gramatoms of Unleached Glass
103
IOO
I0
I
O.l
0.01
0.001
0 2 A .6 .8 I 1.2 1.4
DEPTH (pm)
DEPTH (jLim)
SIMS depth compositional profiles of (a) B; (b) Cs, Sr;
and (c) Fe, U for ABS 118 glass/glass interface after 2-
and 12-month, 90°C burial in Stripa. Data have been
corrected using A1 concentration (adapted from [70]).
Fig. 5-9.

104
increased considerably. However, this interface contains a lot of
alkali(s) within the altered glass surface as compared by the Na and
K profiles in Fig. 5-10 (a). There is no remarkable difference
between the 2- and 12-month spectra for the glass/bentonite
interface. For the glass/granite interface, the 12-month spectrum
shows less attack as compared to that for 2 months, although lower
intensity of the 12-month spectrum suggests a rougher surface as
compared to the 2-month glass sample. The optical micrographs shown
in Fig. 5-8 (d-f) are consistent with the 12-month FT-IRRS data
showing the order of decreasing roughness is glass/bentonite,
glass/glass, glass/granite.
The SIMS compositional depth profiles of the selected elements
for 2- and 12-month ABS 118 glass/glass interface are shown in Fig.
5-9. The 31-month data for ABS 39 and 41 are shown in Fig. 5-4. As
estimated from the B extraction (Fig. 5-9 (a)), the leached layer
thickness of ABS 118 changed from 0.65 pm after 2-month to 1.05 pm
after 12-month burial. Thus, the leach rate of the ABS 118
glass/glass interface between 2- and 12-month burial (0.48 pm/yr) was
smaller by a factor of 8X as compared to that during the first 2
months (3.9 pm/yr). One important feature observed in this figure is
that B concentration dropped from 21 at % to <1 at % within 0.1 pm
thickness and that at both the "outer region" and the "gel" mid¬
plateau region, B is almost completely depleted (also see Table
5-2). The shoulders of Cs and Sr in the 2- and 12-month profiles
(Fig. 5-9 (b)) are located nearly at the identical depth as the B
0
0
o «
or
0
00
c®
c:
E„
»o
c0
too
K)
Oj
00
So Oflt
EE
00
l L
015
100
0 ü
01
««
o o 001
EE
2?
oo
0.001

Gramatoms Remaining Based on IOO Gramatoms Remaining Based on 100
Gramatoms of Unleached Glass Gramatoms of Unleached Glass
105
Fig. 5-10. SIMS depth compositional profiles of (a) Si, H, Na, Li,
K; (b) LD (including La, Ce, Pr, Nd and Y), P, Sn; (c)
Ca, Zn, Ba; and (d) Zr, Mo, Ni, Cr, Si for ABS 118
glass/glass interface after 12-month, 90°C burial in
Stripa.

106
shoulders. However, there are not any shoulders observed in the Fe
and U profiles (Fig. 5-9 (c)), suggesting that these two species are
not very mobile in the glass.
As discussed in Chapter I, release of the elements such as Cs,
Sr and U are the major concern since they undergo spontaneous decay
and emit radioactivity. In this dissertation, the leach depths of B
are used to estimate the radioactivity release. Such estimations do
not take into consideration the actual levels of remaining
radioactive elements as suggested by SIMS profiles of the altered
layers. The estimations, therefore, represent the worst case of the
fission product and actinide release.
Figure 5-10 shows the SIMS profiles of other elements of
interest for the ABS 118 glass/glass interface after 12-month, 90°C
burial. In Fig. 5-10 (a) for Group I elements, Li has been depleted
completely while a considerable amount of Na still remains in the
altered layer. Also observed was an enrichment of K probably due to
the exchange of Na and Li from the glass with K from the ground
water. The large difference in size between K and H would make the
K/H ion exchange less probable. The H buildup at the surface layer
may result from H/alkali exchange and/or hydrolysis of the glass
component oxides. For Group IIA and IIB elements, the profiles (Fig
5-10 (c)) show that Ca and Ba were depleted from the 1.05 pm thick
surface layer while very little depletion of Zn was found at the
surface. The concentration of Si appears to remain almost constant
(see Fig. 5-10 (b)).

107
The FT-IRRS spectra of ABS 118 glass/Pb, glass/Cu and glass/Ti
after 2- and 12-month 90°C Stripa burial are shown in Fig. 5-11. As
compared with the preburial spectrum, very little change in the shape
of the spectra was observed for the glass/Pb interface for up to 12
months. First an increase then a decrease in the specular intensity
for this interface indicates that the roughness of the leached glass
surface first changed to a smooth one then a slightly rough one. But
the thickness of this altered layer appears to be less than 0.5 um.
A greater extent of selective leaching, such as for B and
alkalis, was observed at the glass/Cu and glass/Ti interfaces. As a
result, the roughness of these glass surfaces increased. The
relative SiOg enrichment due to B and alkali depletion results in a
splitting of the broad peak containing the Si-O-Si stretching (S)
vibrations and Si-O-alkali stretching (NS) vibrations. However,
depletion of alkalis appears to be incomplete even after 12 months
burial for these two interfaces.
The optical micrographs shown in Fig. 5-12 confirm the surface
morphology developed during the 2- and 12-month burial as discussed
earlier. In all cases, corrosion products can be found at the glass
surfaces even after short times of exposure.
The SIMS depth profiles of 23 elements as obtained at interfaces
against Pb, Cu and Ti are shown in Figs. 5—13 through 5-19. Table
5-2 lists the averaged gram«atoms of each element remaining at "gel"
mid-plateau and outer region of the three glass interfaces based on
100 gram«atoms of unleached glasses. The SIMS data shown here were

WAVENUMBER (cm")
Fig. 5-11. FT-IRRS analysis of ABS 118 glass/Pb, glass/Ti and glass/Cu interfaces after (a) 2-month
and (b) 12-month 90°C burial in Stripa.
108

109
Fig. 5~12. Light micrographs of ABS 118 glass surfaces after 90°C
Stripa burial for 2 months, (a) glass/Ti, (b) glass/Cu,
and (c) glass/Pb interfaces, and for 12 months, (d)
glass/Ti, (e) glass/Cu, and (f) glass/Pb interfaces.
L

110
the ones corrected by A1 concentration, assuming that the gram*atoms
of this element remain unchanged throughout the leached layer. These
provide data which account for porosities and adsorbed elements. In
Figs. 5-13 through 5-15, the 2- and 12-month profiles of 6 elements
are compared. As shown by B extraction, the leach depth increased
from 0.05 pm to 0.4 pm for the glass/Pb, and from 0.8 pm to 0.85 pm
for glass/Cu between 2 and 12 months and from 0.85 pm to 2.3 pm for
glass/Ti between 7 and 12 months.
All the glass surfaces except the glass/Pb interface show a
complete depletion of B at the leach layer (Fig. 5-13). Cs and Sr
leached nearly to the same depths as B (Fig. 5-14). However, Fe and
U remained unchanged, except at the -0.2 pm outer region where slight
amounts of Fe and U have been dissolved into the ground water during
burial (Fig. 5-15).
Figures 5-16 through 5-19 compare other elemental profiles of
the 12-month glass surfaces for the three interfaces. Sodium, Li,
Ca, Zn and Ba were depleted nearly to the same depths as B. The
decrease in Si concentration at the outer surface of the glass/Pb
interface (Fig. 5-16) is probably due to an enrichment of K and Pb
(Figs. 5-16 and 5-18). It has been noted that almost all of the Li
has been leached out of the glass surface while a considerable amount
of Na still remains at the altered layer, especially for the glass/Cu
case where Na concentration first decreased then increased nearly to
the same concentration as that of bulk glass and finally decreased
again at the outer surface region (<0.2 pm). There are not any

Fig. 5-13. Boron profiles of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 2- and 12-month, 90°C
burial at Stripa.
o
b
o
Gramatoms
Gramatoms
o
o
Remaining Based on I00
of Unleached Glass
Gramatoms Remaining Based on IOO
0
b
Gramatoms
O
of
Unleached Glass
0
0
O
_ 0
Gramatoms Remaining Based on IOO
o Gramatoms of Unleached Glass
IOO

Gramatoms Remaining Based on IOO
Gramatoms of Unleached Glass
112
DEPTH (pm) DEPTH (pm)
DEPTH (pm)
Fig. 5-14. Cs and Sr profiles of ABS 118 (a) glass/Pb, (b) glass/Cu
and (c) glass/Ti interfaces after 2- and 12-month, 90°C
burial at Stripa.

burial at Stripa.
*—*•
Oq
U1
I
LTI
Oí 'D
3 CD
a
Oí
^ 3
o a
a
Oq
H-* *0
Oí T
CO O
CO *-*
\ M-
H M
(D
CO
I—*•
3 O
cf
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*-b CD
Oí CO
O
CD -k
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00
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CT
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3 o'
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cf OQ
O' M
- Oí
CO
CO
o \
o o
o c
o
o
o
Gramatoms Remaining Based on 100
Gramatoms of Unleactied Glass
o
100
DEPTH (pm) DEPTH (pm)
o
b
o
Gramatoms Remaining Based on 100
Gramatoms of Unleached Glass
O
o
b
o
Gramatoms Remaining Based on 100
Gramatoms of Unleached Glass
o
2 ° -
001 1 * 001

Gramaioms Remaining Based an 'OG
Gramatoms of Unleached Glass
114
DEPTH (jum)
Fig. 5-16. SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90°C Stripa
burial, Si, H, Li, Na and K profiles.

Gramatoms Remaining Based on iOO
Gramaroms of Unleached Glass
115
DEPTH (,um)
DEPTH (jam)
Fig. 5-17. SIMS analysis of ABS 118 (a) glass/Pb, (b) glass/Cu and
(c) glass/Ti interfaces after 12-month, 90°C Stripa
burial, Ca, Zn and Ba profiles.

116
obvious changes in the concentration of LD (including La, Ce, Pr, Nd
and Y), Zr, Mo and Cr in the leached layer except at the -0.2 ym
outer region where a decrease in the concentration of these elements
can be observed (Figs. 5-18 and 5-19). However, an increase in the
Ni concentration was apparent at the "gel" mid-plateau region of the
glass/Cu and glass/Ti interfaces after 12-month, 90°C burial.
It is noted that there is an apparent Pb buildup at the glass/Pb
and an increase of Cu concentration at the glass/Cu interfaces (Table
5-2 and Fig. 5-18). A large peak including Pb and U in the RBS
analysis for the glass/Pb interface (Fig. 5-20) confirms the buildup
of Pb at this interface after 12-months of burial. The maximum mass
(1 gram*atoms) of Pb was located at -0.2 ym below the glass surface
in the case of glass/Pb (Fig. 5-18 (a)) while the Cu concentration
reached its maximum value at -0.9 ym of the leach glass surface
exposed to Cu during burial (Fig. 5-18 (b)).
Results with SRL Glasses
Figure 5-21 shows the FT-IRRS analysis of the glass/glass
interface for three SRL glasses after 2-year burial in Stripa at
8-10°C and 90°C. The 90°C spectra suggest the depletion of alkalis
and boron accompanied by the alteration of the Si02 network within
-0.5 ym detection limit of FT-IRRS. Also shown in this figure is
that the most extensive leaching occurs on SRL 131 + 29.8% TDS glass
among the three SRL glasses, after 2-year burial. The spectrum of
this glass surface has the lowest peak intensity and only one main
peak in the 800-1100 cm-1 region. As regards the low temperature
(8-10°C) spectrum for SRL 131 + 29.8% TDS glass/glass interface, the

'TI
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Gramatoms Remaining Based on 100
Gramatoms of Unleached Glass
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Gramatoms Remaining Based on 100
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Gramatoms
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117

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Gramatoms Remaining Based on 100
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DEPTH (pm) DEPTH (pm)
Gramatoms Remaining Based on 100
Gramatoms of Unleached Glass
o
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Gramatoms Remaining Based on 100
Gramatoms of Unleached Glass
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Table 5-2. GranrAtoms of Elements Remaining at Gel Mid-Plateau and Outer Region of the Altered
Glass Surface Based on 100 Gram*Atoms of Unleached ABS 118 glass E after 12-Month,
90°C Burial in Stripa.
Gel Mid-Plateau
Outer Region
Glass/Pb Glass/Cu Glass/Ti
Glass/Pb Glass/Cu Glass/Ti
Si
110
23
43
41
24
44
41
Li
6
2
0.07
0.5
1
0.05
0.5
Na
16
9
14
3
7
11
3
K
0.05
0.8
0.6
0.14
0.7
0.3
0.11
Cs
0.4
0.1
0.5
0.14
0.04
0.3
0.11
Mg
0.05
0.9
0.04
0.05
0.7
0.06
0.09
Ca
4
0.8
0.8
0.7
0.3
0.6
0.6
Sr
0.15
0.02
0.08
0.03
0.007
0.06
0.02
Ba
0.2
0.02
0.09
0.1
0.004
0.07
0.1
Zn
0.5
0.2
0.2
0.4
0.1
0.1
0.3
B
21
5
0.7
0.4
2
0.5
0.1
A1
5.4
5.4
5.4
5.4
5.4
5.4
5.4
Mn
0.5
0.1
0.09
0.4
0.08
0.08
0.1
Fe
2
1 .1
2
2.1
0.8
1.4
2.1
Zr
1
0.2
0.9
0.9
0.1
0.8
0.8
Mo
0.7
0.1
0.8
0.8
0.06
0.6
0.6
LDa
1
0.1
0.9
0.9
0.04
0.5
0.6
U
0.16
0.02
0.1
0.1
0.005
0.08
0.09
Pb
(0.007)
1
-
-
0.6
-
-
Cu
(0.007)
-
0.2
-
-
0.08
-
Ti
(0.07)
-
-
0.04
-
-
0.03
x b
pm
(0.001 )
0.4
0.9
2.3
a "LD" stands for the sum of La, Ce, Pr, Nd and Y.
b x(jm: Approx, depth of leached layer.
119

INTENSITY (Arbitrary Units)
1 20
Fig. 5-20. RBS analysis of ABS 118 glass/Pb, glass/Ti and glass/Cu
interfaces after 12-month, 90°C burial in Stripa.

Fig. 5-21.
FT-IRRS analysis of the glass/glass interface for three SRL glasses after 2 years of burial
in Stripa.
121

intensity of the main peak containing Si-O-Si stretching vibrations
decreased by only 3? as compared with a 12? (4X) decrease for the
same glass buried at 90°C.
SEM micrographs of the glass/glass interface for the three SRL
glasses after 2-year burial at 90°C in Stripa are shown in Fig.
5-22. These pictures provided representative views of these
surfaces. Polishing scratches are clearly visible on the preburial
surface (Fig. 5-22 (d)). Figure 5-22 (b) shows the vestiges of the
original polishing scratches and pits on a leached SRL 165 + 29.8?
TDS glass/glass interface indicating very little surface attack. On
SRL 131 glasses with either 29.8? TDS or 35? TDS waste, the original
polishing scratches have been removed through leaching (Fig. 5-22
(a, c)).
As revealed by SIMS analysis (Fig. 5-23 (a-c)), the leach depths
of these three glasses are 3.6 ym for SRL 131 + 29.8? TDS, 0.46 pm
for SRL 165 + 29.8? TDS and 1.85 pm for SRL 131 + 35? TDS after
2-year, 90°C burial. These data indicate that SRL 165 frit is much
better than SRL 131 and that an increase in waste loading from 29.8
wt? to 35 wt? is also beneficial (i.e., decreased the leach depth of
SRL 131 glasses by 2X).
The common feature observed from the FT-IRRS and SIMS analyses
is preferential leaching of alkalis and boron contained originally in
the glasses. Although some dissolution of Si is observed from the
SIMS compositional profiles as shown by the decrease in its
gram-atoms remaining based on 100 gram-atoms of unleached glass, the

123
Fig. 5-22. SEM micrographs of glass surfaces in contact with glass
of the same composition during 2-year burial at 90°C in
Stripa: (a) SRL 131 + 29.8% TDS, (b) SRL 165 + 29.8?
TDS, (c) SRL 131 + 35? TDS and (d) an uncorroded glass
surface.

Fig. 5-23. SIMS in-depth profiles of glass surfaces after 2-year
Stripa burial at 90°C.
GRAM ATOMS REMAINING BASED ON
100 GRAM ATOMS OF UNLEACHED GLASS
O
O
O
o
8
GRAM ATOMS REMAINING BASED ON
100 GRAM ATOMS OF UNLEACHED GLASS
O
O
O
O
b
o
_ o
o o
GRAM ATOMS REMAINING BASED ON
100 GRAM ATOMS OF UNLEACHED GLASS
100

125
amount of dissolved Si is small. This is mainly due to (1) that SÍO2
is the glass former and the Si-0 bonds are strong (106
kcal/gram*atom) and (2) the buffering effects of the ground water in
the Stripa mine where pH showed less than 1 unit increase over its
original value of 8.1 during the whole span of burial.
Effect of Glass Heterogeneities
The x-ray diffraction (XRD) pattern in Fig. 5-24 shows that some
SRL 131 + 29.8% TDS glass samples contained Fe^O^ spinel crystallites
[73]* The peak at 20 = 33-1° is due to the Fe20^ contamination from
the iron mortar during sample preparation [74]. The crystalline
phase is a spinel solid solution, (Mn, Fe, Ni)1 Fe20i|, containing
small amounts of Mn and Ni, as indicated by EDS analyses for the
crystal areas of a preleached heterogeneous glass shown in Fig. 5-25
(b).
The SEM-EDS analysis for SRL 131 + 29.8% TDS glass is shown in
Fig. 5-25 for preleached specimens both with and without crystals.
In general, the surface of the glass appears as that shown in Fig.
5-25 (a). This micrograph reveals a homogeneous surface with a few
polishing scratches. Some areas of the glass surface appear as shown
in Fig. 5-25 (b). The microstructure observed in these regions
consists of an isolated crystalline phase within a glassy matrix.
The specimens contain less than 20% crystals which are nonuniformly
dispersed within the glass matrix. This value was estimated using
point counts from optical micrographs obtained from a number of
representative sections. Energy dispersive spectra for the
crystalline area of the heterogeneous specimens are shown in

INTENSITY
29
Fig. 5-24.
X-ray diffraction pattern for powders prepared from devitrified SRL 131 + 29.8% TDS glass.
1 26

127
KEV
(b)
Fig. 5-25. SEM-EDS analysis of preburial SRL 131 + 29.8% TDS
glass: (a) homogeneous glass surface and (b) partially
devitrified glass surface.

Fig. 5-25 (b). These data reveal that the crystalline phase is rich
in Fe, Ni and Mn. The presence of the Si peak is probably due to the
silicate phase since the crystals are too small for isolated analysis
without interference from the surrounding glass.
Uniform corrosion was observed on the homogeneous specimens
buried for up to 6 months at 90°C. Figure 5-26 is the SEM of the
glass surface interfaced with bentonite after 1-month burial at
90°C. This, together with EDS data in Fig. 5-28 and optical
micrographs, suggests the formation of an uniformly leached surface
layer.
The leaching morphology of the samples in the devitrified areas
(or areas containing undissolved raw materials) appears nonuniform.
The SEMS for these glass specimens after 1-, 3~ and 6-month burial at
90°C are shown in Fig. 5-26 (b, c, d), respectively. The surfaces
have sharply projecting crystals and deep furrows at the crystal-
glass interface. These micrographs suggest that the crystalline and
parent glass phases are more resistant to aqueous attack than are the
interfacial regions. The extent of phase boundary attack increases
during the first month of burial, but shows very little change
between 3 months and 6 months. An additional feature observed on the
6-month sample is cracking of the glassy phase between the
crystalline phases. The cracking is indicative of extensive leaching
in these regions. Leaching occurs in the similar regions of the 1-
and 3-month samples, but is apparently not extensive enough to cause
cracking. Another feature shown in these micrographs is pitting,

129
Fig. 5-26. SEM micrographs of SRL 131 + 29.8% TDS glass surfaces in
contact with bentonite, Stripa burial at 90°C: (a)
homogeneous glass, 1-month burial; (b) partially
devitrified glass, 1-month burial; (c) partially
devitrified glass, 3“month burial; and (d) partially
devitrified glass, 6-month burial.

130
which can be seen easily within the crystalline areas for all the
heterogeneous specimens buried for 1, 3 and 6 months at 90°C. These
observations are consistent with the FT-IRRS spectra which show
decreasing peak intensities and reduced integrated areas (Fig. 5-27
(b-d)) for longer exposure times.
The FT-IRRS spectra for homogeneous and heterogeneous glass
specimens both prior to and after burial are shown in Fig. 5-27.
There is comparably less change in the FT-IRRS integrated area for
the homogeneous glass specimens after burial at 90°C for 1 month than
for the heterogeneous glass specimens. It is also worth noting that
a fine structure appears in the FT-IRRS spectra below 700 cm”1 for
the latter specimens after 1-, 3~ and 6-month burial at 90°C. Two
peaks, at 602 and 530 cm”1, were identified as the characteristic
peaks for trevorite, NiFe20ij. This is consistent with the XRD [75]
and EDS data showing a Fe- and Ni-rich crystal phase in the
heterogeneous glass specimens. The enhancement of the fine structure
of the FT-IRRS spectra due to trevorite is a consequence of a higher
leach resistance of the crystalline phase than that of the glassy
matrix. As a result, the crystalline phase projects out of the glass
surface as time increases resulting in more of a contribution to the
FT-IRRS spectra from the crystalline phase. This provides evidence
for a less durable glassy matrix (as compared with the crystalline
phase) and a preferential phase boundary attack of the heterogeneous
samples during burial.

5/o Reflectance % Reflectance
131
Wavenumbers (cm1) Wavenumbers (cm1)
(a)
(b)
(c)
(d)
Fig. 5-27. FT-IRRS analysis of SRL 131 + 29.8? TDS glass surfaces in
contact with bentonite, Stripa burial at 90°C: (a)
homogeneous glass, 1-month burial; and partially
devitrified glass (b) 1-month burial; (c) 3-month burial;
and (d) 6-month burial.

The EDS data for homogeneous glass specimens prior to burial is
shown in Fig. 5-28 (a). It was observed that the Fe:Si peak ratio
increased from 1:4.7 for the uncorroded specimen (Fig. 5-25 (a)) to
1:3.7 for the burial specimen interfaced with bentonite at 90°C for 1
month. This suggests that the layer formed on the glass surface
interfaced with bentonite at 90°C for 1 month burial is Fe-rich. In
the case of the heterogeneous samples, the EDS data show even more Fe
enrichment within the surface layer of the glassy phase (Fig. 5-28
(b)). In this case, the Fe:Si peak ratio reaches 1:3.2 indicating
that crystallization favors Fe enrichment during burial. With
increasing burial times, EDS data (Fig. 5-28 (c, d)) exhibit an
increase in Fe:Si peak ratios (1:3-2 for 1 month, 1:1.4 for 3 months
and 1:0.9 for 6 months).
The EDS data show that for uncorroded specimens, the main
constituents of the crystal phase are Fe202, NiO and MnO and possibly
some Si02 (Fig. 5-25 (b)). This suggests a compositional variation
of spinel solid solutions. For 3“ and 6-month specimens, a small
increase in Fe:Ni peak ratio can be observed (Fig. 5-29 (b, c)).
These data indicate for the crystalline areas an Fe:Ni ratio of
approximate 8.8:1 for 1-month, 9.5:1 for 3-month and 10:1 for 6-month
burial specimens. This increase in Fe:Ni peak ratios is much smaller
than the change in Fe:Si peak ratio for the glassy matrix.
Alkali ions, such as Na, are leached from the glassy surface
during burial. This may be assessed from the EDS data which show
lower Na peak intensities for homogeneous glass specimens buried at

Relative Intensity Relative Intensity
133
(a)
(h )
kEV
kEV
(c)
(d)
Fig. 5-28. EDS analysis of SRL 131 + 29.8? TDS glass surfaces in
contact with bentonite, Stripa burial at 90°C: (a)
homogeneous glass, 1-month burial; and glass matrix of
partially devitrified glass, (b) 1-month burial; (c)
3-month burial; and (d) 6-month burial.

Relative Intensity
131»
(b)
... .i ... â– â– 
0 2.6 5.1 7.7
kEV
(C)
Fig. 5-29. EDS analysis of crystal areas of partially devitrified
SRL 131 + 29.8? TDS glass surfaces in contact with
bentonite, Stripa burial at 90°C: (a) for 1 month, (b)
for 3 months and (c) for 6 months.

135
90°C for 1 month, and the disappearance of the Na peak for the glass
matrix areas of the heterogeneous glass specimens interfaced with
bentonite after 1-, 3~ and 6-month burial at 90°C. On the other
hand, the compositional change in the crystalline phase seems
relatively small, the only exception being Si, which shows an obvious
relative depletion after burial (Fig. 5-29). The Si peak shown in
the spectra of the heterogeneous samples is most likely due to
contributions from the surrounding glassy phase.
Laboratory Test Results
Modified MCC-1 Static Leach Tests
Field tests were simulated in the laboratory by changing three
variables. First, water from the Stripa mine was used to replace the
deionized water typically used in laboratory experiments. Secondly,
the SA/V was increased from 0.1 cm-1 (typical in lab tests) to 1.0
cm 1. Although this latter value is still small compared to the
SA/Vs obtained during burial, it represents a practical working value
above which it is difficult to control in the laboratory when bulk
glass samples are used. Thirdly, a "saturated" Stripa water was
prepared by exposing regular Stripa water to a large SA/V using glass
powders.
Figure 5~30 shows the FT-IRRS results of the laboratory tests
for SRL 165 + 29.8% TDS glass. Three conclusions can be drawn from
these results: (1) higher SA/V ratios result in less surface
deterioration; (2) the extent of surface deterioration is less in
ground water than in deionized water; and (3) water that already

Fig. 5-30. FT-IRRS analysis of SRL 165 + 29.8% TDS glass before and after leaching for 28 days at 9Q°C
in (a) deionized water with SA/V = 0.1 cm , (b) Stripa ground water with SA/V = 0.1 cm ,
(c) Stripa ground water with SA/V = 1.0 cm and (d) saturated Stripa ground water with
SA/V = 0.1 cm” . Also shown is a spectrum for the glass/glass interface after 3-month
Stripa burial.

WAVENUMBERS

contains large concentrations of glass species, i.e., "saturated"
Stripa water, is less aggressive than regular ground water or
deionized water. The major factor contributing to the reduction in
leaching is the presence of Si in solution. Si is not present in
deionized water, but Stripa water contains about 13 ppm of Si and
"saturated" Stripa water contains 85 ppm. As shown in Fig. 5~30,
when other conditions are equivalent, including SA/V, the reduction
in the FT-IRRS peak intensity is inversely related to the Si
concentration in the laboratory tests. Increasing the SA/V from 0.1
cm 1 to 1.0 cm 1, while maintaining other conditions constant,
results in a decrease in the change in peak intensity. Again, this
is related to the Si concentration, although the initial Si
concentration is the same in both tests, i.e., when SA/V
= 0.1 cm \ the Si concentration increases more rapidly in the
solution than with SA/V = 1.0 cm-1. Additionally, less Si is
required to be released from the glass when SA/V = 1.0 cm 1 in order
to approach saturation. Therefore, less surface deterioration is
expected when SA/V = 1.0 cm-1, as is confirmed by the spectra in Fig.
5-30.
SIMS analysis of SRL 165 + 29.8% TDS glass leached in the
laboratory under conditions of 1 month, 90°C and SA/V = 1.0 cm 1 is
shown in Fig. 5—31 - From this figure, it can be seen that B, Na, Li
and U are depleted in these laboratory-leached specimens. A small
amount of Si was also leached out, Ca was enriched and Fe remained
almost unchanged at the surface. The depth of leaching on the glass
represented in Fig. 5—31 is about 0.5 pm.

GRAM ATOMS REMAINING BASED ON
139
Fig* 5-31 * SIMS analysis of SRL 165 + 29.8% TDS waste, laboratory-
corroded, 1 month 90°C in Stripa water with SA/V = 1.0
cm

Single-Pass Flow Tests and Static Tests Using Rock Cups
Figure 5-32 is a plot of solution pH vs time for Black Frit 165-
Mobay glass corroded under static and flow test (0.3 mL/h)
conditions. The leachant was collected each week for up to 26 weeks
during the flow tests and pH was measured after each collection. The
original ground water pH was 7.6 ± 0.1.
In the case of flow tests, the pH changed by 0.8 unit (from 7.3
to 8.1). The pH vs time curve (Fig. 5-32) shows an initial small
increase at 9-10 weeks with a maximum of 8.1, and then buffering at
slightly lower pH values.
In contrast, the static leachant exhibited higher pHs (7.5-
8.4). This is not surprising since under static leach conditions,
the alkali ions which were leached out from the glass surface
accumulated in the solution and resulted in higher pH values.
However, this pH increase was small.
The ground water buffering has a strong action on glass
leaching, since dissolution of the glass formers, Si02 and B^^, is
highly pH-dependent. Below pH 9, the activity of Si02 is low
(<10“4 M for vitreous silica at 25°C [76] and slightly higher for
borosilicate glasses). Thus, a skeleton rich in Si02 could be
preserved.
The concentrations of Si, B, A1 and Li in the single pass
flowing ground water as a function of leaching time, are shown in
Fig. 5-33. An important feature observed in this figure is an
initial increase in the solution concentration to a maximum followed
by a decrease with time. Two possible reasons for this are (1) the

Fig. 5-32.
Solution pH vs time for Black Frit
conditions.
corroded under static and flow

CONCENTRATION (mg/L)
1 42
TIME (weeks)
Fig. 5-33. Concentrations of Si, B, A1 and Li in the single pass
flowing ground water at 0.3 mL/hr as a function of
leaching time for Black Frit 165-Mobay glass samples.

higher dissolution rate of edges of polishing scratches [77] and (2)
as suggested by Grambow [47] a diffusion barrier established after
4-5 weeks leaching, capable of slowing down the elemental release
from the inner glass.
The selective leaching of Li and B is quite visible in Fig. 5—34
where the normalized leach rates of Li, B and Si, and weight loss are
plotted vs leach time. In this figure, both flow and static test
results are shown. In the case of flow tests, the normalized mass
loss of Li and B are larger than that of Si. Also, Li leached faster
than B.
Figure 5~35 shows the EDS analysis of the Frit 165-Mobay glass
surfaces prior to and after static leaching at 90°C. The peak
intensity ratios to Si for various elements in the glass are listed
in Table 5-3. These data confirmed the preferential leaching of
alkalis as shown by the decrease in the Na to Si peak ratio vs
time. The A1 to Si peak ratio remained almost unchanged and the Fe
to Si peak ratio increased from 0.11 to 0.16 after 6-month static
leaching. These data suggest that the least-soluble species, such as
Fe and Al, stay in the leached layer.
The FT-IRRS and SIMS data for Frit 165-Mobay glass leached in
the granite cup tests at 90°C under static and flow conditions are
shown in Figs. 5-36 and 5—37. As shown in Fig. 5-36, surface
smoothening is revealed by an increase in the spectral intensity
after 1-month leaching at 90°C as compared with an uncorroded
surface. This is because the sharp edges of the polishing scratches
have smaller curvatures and higher dissolution rates as compared to a

1 44
30
Black Frit 165-Mobay
Granite Cup Tests
. 90° C, SA/V *1.0 cm1
Flow at 0.3ml /hr:
°V25
/ x - Li
/ v -B
/ • -Si
” / ■ - Weight loss
\
/ Static'
O'
f • - Si
/ 0 - Weight loss
0>
f j/7b
o20
/
tr
•0
0 I5
l /

i
/ 7 >r
•0
Nl
0 I0
E
k-
O
5
]!/ -
uy-ja— i 1 i zrj i
0 5 10 15 20 25 30
O'
c 5
5 £
Time (weeks)
Fig. 5-34. Normalized leach rates of Li, B and Si as a function of
time under flowing (at 0.3 ml/hr) conditions for Black
Frit 165-Mobay glass with SA/V = 1.0 cm . Also shown
are the weight losses for glass samples leached under
static and flow conditions with SA/V = 1.0 cm

Relative Intensily Relative Intensity
1 45
2.6 5.1 7.7
KEV
Fig* 5~35. EDS analysis of Black Frit 165-Mobay glass leached in the
rock cup test at 90°C with SA/V = 1.0 cm in ground
water under static conditions: (a) uncorroded; (b) for 1
month; (c) for 3 months; and (d) for 6 months.

Relative Intensity Relative Intensity
1 it 6
BlacK Fir it 165 -Mobay
Granite Cup Static Test
90° C , 1 cm'1, 3 months
s
i
Al
Fe
NaJJ
V Ca Mn fy pe
2.6 5.1 7.6
KEV
Figure 5-35.—continued

Table 5-3. Relative Concentrations (Ratio to Si) at the Black Frit 165-Mobay Glass
Surface After Static Leaching in the Rock Cup Test. Data Are from EDS
Analysis.
Leach
Time
Fe
Na
A1
Ca
Ti
Mn
Ni
Mg
Unc.
0.11
0.049
0.096
0.043
0.017
0.027
0.015
0.026
1 mo.
0.087
0.095
0.109
0.037
0.01 4
0.024
0.010
0.036
3 mo.
0.10
0.044
0.091
0.040
0.015
0.027
0.012
0.027
6 mo.
0.16
0.024
0.097
0.079
0.023
0.039
0.017
0.022
4=-

1200 1000 800 600
WAVENUMBERS
Fig- 5-36. FT-IRRS analysis of Black Frit 165-Mobay glass leached in the granite rock cup test at 90°C
under static conditions with SA/V = 1.0 cm
8ft L

flat glass surface [77]. After 3 months, selective leaching of
alkalis and boron (Fig. 5~37 (a)) caused splitting of the peak at
800-1200 cm-1 split and resulted in a spectral intensity decrease at
the same region (see Fig. 5-36). The peak position of the Si-O-Si
stretching vibrations shifted to a higher wave number as a result of
Si02 enrichment within the leached glass surface. Leaching of alkali
and boron also resulted in a decrease in the peak intensity of the
silicon-oxygen-alkali stretching vibrations. The drop of the peak
intensity for the Si-O-Si stretching vibrations is due to surface
roughening through selective leaching and network dissolution after 3
months.
After 6 months, a further decrease in the spectral intensity
within the region of 800-1200 cm-1 is observed (Fig. 5-36). This
roughening is basically due to further dissolution of Si02, the major
network formers, selective leaching of B20^ and alkalis. Most
species, such as Fe-^O^ and A^O^, exhibit extremely low solubility
limits in the ground water with pH = 7-8 and thus stay at the glass-
water interface or deposit on the walls of the corrosion pits which
further roughened the glass surface (Fig. 5—37 (c)). The additional
enrichment of Si02 at the altered glass surface layer is indicated by
shifting of the Si-O-Si stretching vibration peak to the higher wave
numbers. Since leaching of alkalis and boron was more complete after
6 months (Fig. 5—37 (c)), the intensity of the silicon-oxygen-alkali
peak decreased (Fig. 5~36). Also, separation of the two peaks became
more obvious. These findings are consistent with other analytical
data as discussed earlier.

GRAM-ATOMS REMAING BASED ON
IOO GRAM-ATOMS OF UNLEACHED GLASS
DEPTH (pm)
(a)
Fig. 5-37. SIMS depth profiles for Frit 165-Mobay glass leached in the granite rock cup tests at 90°C
with SA/V = 1.0 cm-1, (a) under static conditions and (b) under flow conditions (0.3
mL/hr).
150

DEPTH(pm)
GRAM ATOMS REMAINING BASED ON
100 GRAM ATOMS OF UNLEACHED GLASS
o
100

CHAPTER VI
DISCUSSION
ABS Glasses
As shown by the FT-IRRS spectra, optical micrographs and SIMS
depth profiles in Chapter V, the two glasses, ABS 39 and ABS 41,
displayed a considerable difference in leachability during 90°C,
31-month Stripa burial. An earlier laboratory study [17— 19U
indicated that ABS 41 glass was three to four times more durable than
glass ABS 39. The 31 -month burial data confirm that glass ABS 41 is
markedly superior to glass ABS 39 with respect to attack of
glass/glass and glass/granite interfaces. Figure 6-1 summarizes the
time dependence of the thickness of the leach layer based on boron
depletion for both glasses at the glass/glass, glass/granite and
glass/bentonite interfaces. As shown in Fig. 6-1 for glass ABS 41,
the effect of bentonite is significant. The leached layer at this
interface is 10 to 20 times thicker than those of the glass/glass and
glass/granite. For this glass, the leach rate, as estimated by the
slopes of the curves, was slowed down greatly after the first 1 to 3
months of burial, but the subsequent decrease in leaching rate is
seen to be very moderate, if any. For example, the leachability at
the ABS 41 glass/bentonite interface was 1.4 pm/yr during the first
month of burial, 7.2 pm/yr between 1 to 3 months, but 12 pm/yr
152

LEACH DEPTH (jjm) LEACH DEPTH (um)
153
(b)
/
/
/
/
/
/
/
/
/
I0-
/
ABS 4I
STRIPA BURIAL
3I Mo., 90° C
Glass/ Bentonite
/
/
r Glass/Granite
r Glass/Glass
*—=
3I
BURIAL TIME (Months)
Fig. 6-1. Time dependence of reaction layer thickness for glass ABS
39 (a) and ABS 41 (b) after 31“month, 90°C Stripa burial.

between 3 to 31 months, based upon depth of boron removal. In
contrast, the teachability of the ABS 41 glass/glass interface was
0.8 pm/yr during the first 12 months and 0.03 ym/yr for 12 to 31
months of 90°C burial. The teachability of the ABS 41 glass/granite
interface was 0.4 pm/yr within the first 12 months and approximately
0.24 pm/yr between 12 and 31 months of 90°C burial.
Thus, during the thermal period of storage (-300 years), when in
contact with glass of the same composition under Stripa burial
conditions, ABS 41 will leach to a depth of less than 9 pm, even if
exposed to water immediately after burial. This can be compared to
layer thicknesses of the order of 2,700 pm for the glass/glass
interfaces of ABS 39 (Table 6-1 and Fig. 6-1 (a)) during the same
period.
For ABS 39, the presence of bentonite apparently did not
appreciably accelerate glass leaching after the initial 3 months.
This is a significant finding since, up to 12 months, the ABS
glass/bentonite interfaces leached four times faster than the
glass/glass interfaces. It is the increased leach rates at both
glass/glass and glass/granite interfaces between 12 and 31 months
that reduced the difference in the leach depths between these
interfaces. One possible explanation is a greater exposure to water
at the ABS 39 glass/glass and glass/granite interfaces. The Stripa
underground laboratory is believed to provide a relevant repository
environment with high SA/V ratios and low ground water flow rates.
However, the actual amount of water which enters or passes through a
particular interface may vary locally. Therefore, there could be an

155
Table 6-1. 90°C Glass Leach Rates During 12- to 31-Month Period
(pm/yr).
Glass/Glass
Glass/Granite Glass/Bentonite
ABS 39 9
19
1 .3
ABS 41
0.03
0.24
12

increase in the teachability of the glass/glass and glass/granite
interfaces if the ground water flow rate increased and/or the SA/V
ratio decreased locally. The local attack aspect, as mentioned
above, is also likely to be of importance for the accelerated
leaching at prolonged times, especially at the glass/granite
interfaces.
The optical micrographs indicate that this might have been the
case. The rather thick and porous surface films formed on these two
interfaces are evidence that these glasses leached at a low SA/V
ratio or a higher ground water flow. This kind of surface film is
characteristic of glass surfaces after an MCC-1 static test with SA/V
= 0.1 cm-1 or an MCC-4 test with a flow rate >0.5 mL/h [78].
Based on the plot of leach depth vs burial time for the ABS 39
glass/bentonite interface, the leach rate was 5.6 pm/yr during the
first 3 months of burial and 1.3 pm/yr between the 3rd and 31st
month. Thus, the leach depth after the thermal period of storage
('300 years) can be estimated to be of the order of 400 pm for the
ABS 39 glass/bentonite interface.
It has been shown many times, and discussed in a recent
publication [79], that ABS 41 has a considerably smaller leach depth
than ABS 39. This is thought to be due to a higher (Si + Al)/(alkali
+ B) ratio, and perhaps also to the coexistence of sodium and lithium
in the bulk glass, i.e., a "mixed-alkali effect" which may reduce the
rate of ion exchange [80,81]. ALso, it has been shown [79] that the
presence of divalent cations, such as zinc, may further reduce the
rate of leaching through the formation of a protective zinc silicate

157
layer. The low depletion rate for ABS 41 is an important finding
since this composition is quite close to that chosen for commercial
high-level waste solidification by the French for LaHague
operations. However, the observed sensitivity of ABS 41 to long-term
attack at the bentonite interface is disconcerting. One contributing
factor may be the presence of zinc, which, according to other
evidence [79], usually enhances leaching resistance at short times,
but might be less advantageous at longer exposures. This could be
due to a pH effect where bentonite creates high pH environment
dissolving Zn(0H)2 and Zn silicates from the glass surface.
In comparing the leach rates among the three ABS glasses, the
SIMS data shown in Fig. 6-2 and Table 6-2 indicate that both ABS 118
and ABS 41 are superior to ABS 39. The ABS 118 glass/glass and
glass/granite interfaces leached just slightly faster than those of
ABS 41, but the ABS 118 glass/bentonite interface shows a better
leach resistance than ABS 41. For these three glasses, the
glass/granite interfaces leached more slowly than the glass/glass
interfaces.
Table 6-2 lists the SIMS surface compositional analysis of the
glass/glass, glass/bentonite and glass/granite interfaces for the
three ABS glasses after 12-month, 90°C Stripa burial. A common
feature observed from the SIMS analysis is that, within the altered
glass surfaces, almost all of boron has been depleted while a
considerable amount of Na still remains in these glasses. This is
probably due to the presence of Na ions in ground water before burial
(Table 4-6). Thus the release of Na from glass was somewhat

Table 6-2. SIMS Compositional Analysis of Glass/Glass, Glass/Bentonite and
Glass/Granite Interfaces for ABS 39, ABS 41 and ABS 118 after 12-Month,
90°C Stripa Burial (Gram*atoms Remaining Based on 100 Granratoms of
Unleached Glass).
ABS 39 ABS 41
"Gel," Mid-Plateau "Gel," Mid-Plateau
Bulk
Glass/
Glass
Glass/
Bentonite
Glass/
Granite
Bulk
Glass/
Glass
Glass/
Bentonite
Glass/
Granite
Si
40.7
42.27
20.71
25.37
42.1
40.22
16.34
46.15
Li
(0.01)
0.02
9.8
0.94
0.1 7a
0.22
Na
20.85
5.64
2.41
2.74
15.55
6.68
1.69
2.58
K
0.04
0.27
0.32
0.25
0.05
0.24
0.31
0.43
Cs
0.30
0.05
0.02
0.12
0.30
0.15b
0.01b
0.22
Mg
0.02
n .m.c
n.m.
0.27
0.03
0.07d
0.56
0.31
Ca
(0.01 )
0.61
0.71
0.39
0.01
0.24
0.54
0.68
Sr
0.10
0.04
0.02
0.08
0.10
0.04b
0.02
0.04
Ba
0.15
0.01
0.04
0.01
0.15
0.15b
0.003
0.07
Zn
-
-
-
-
1 .80
1 .28
0.09
1 .38
B
27.45
1 .28
0.13
1 .09
22.2
2.98a
0.34a
2.46
A1
3.05
3.05
3.05
3.05
2.4
2.4
2.4
2.4
Mn
0.45
n.m.
n.m.
0.05
0.45
n.m.
n.m.
0.22
Fe
3.60
4.30
2.57d
2.93
1 .80
2.4
1 -75í
2.4
Zr
0.55
0.55
0.59d
0.43
0.55
0.58
0.47d
0.49
Mo
0.80
n.m.
n.m.
0.57
0.75
n.m.
0.01
0.12
Y
0.70
0.04
0.03
0.05
0.06
0.05
0.01
0.04
La
0.20
0.15
0.05
0.07
0.20
0.18
0.02
0.12
U
0.30
0.24
0.03
0.23
0.30
0.21
0.01
0.18
x e
4.0
14.8
1.8
1 .0
3.8
0.36

Table 6-2—continued.
ABS 118
"Gel," Mid-Plateau
"Outer Region"
Glass/ Glass/ Glass/
Glass/
Glass/
Glass/
Bulk
Glass Bentonite Granite
Glass
Bentonite
Granite
Si
40
41
41
34
41
39
33
Li
6
0.02
0.01
0.3
0.02
0.01
0.4
Na
16
8
5
6
9
2.6
6
K
0.05
0.3
0.7
0.5
0.2
0.6
0.6
Cs
0.4
0.2
0.03
0.05
0.2
0.02
0.03
Mg
0.05
0.06
0.3
0.4
0.3
0.4
0.4
Ca
4
1
0.9
0.6
0.9
0.9
0.4
Sr
0.15
0.06
0.007
0.02
0.05
0.006
0.01
Ba
0.2
0.1
0.04
0.01
0.1
0.03
0.01
Zn
0.5
0.4
0.1
0.1
0.5
0.005
0.1
B
21
0.4
0.2
1 .5
0.2
0.06
1
A1
5
5
5
5
5
5
5
Mn
0.5
0.5
0.4
0.04
0.2
0.2
0.02
Fe
2
2.1
2
1.3
13
1
0.8
Zr
1
0.9
1
0.5
0.7
0.7
0.2
Mo
0.7
0.7
0.8
0.4
0.3
0.02
0.2
LDf
1
0.9
0.8
0.3
0.4
0.07
0.1
U
0.16
0.1
0.1
0.05
0.06
0.01
0.01
a
b
c
d
e
f
Concentration increasing with depth in "gel."
Concentration varies, minimum in "gel" zone,
n.m.: Not measured.
Concentration decreasing without "plateau."
X : Approx, depth of leached layer.
"ÃœD" stands for the sum of La, Ce, Pr, Nd and Y.
159

BORON DEPLETION DEPTH (¿jm)
160
IOO
IO
I.O
O.l
I 2 3 7
TIME (Months)
I2
Fig. 6-2. Boron depletion depth vs burial time for the glass/glass,
glass/granite and glass/bentonite interfaces. Three ABS
glasses are compared.

161
suppressed due to a decreased concentration gradient. Data in Table
6-2 also confirm that at the ABS 41 glass/bentonite interface, more
SÍO2 has been dissolved from the "gel*" mid-plateau region and Na
depletion was more complete as compared to that of ABS 118. SIMS
analyses also show that very little leaching of transition metals
such as Zr, Mo, Ni and Cr occurred (Fig. 5-10).
SRL Glasses
The experimental results for the three SRL glasses are
summarized in Fig. 6-3 based on the boron depletion depth from the
SIMS analysis. A general tendency of the time-dependence of the
leach depth can be seen clearly from the figure. After an initial
rapid increase, the leach rates slowed down after 1-3 months of
burial.
There was an appreciable effect of composition on SRL waste
glass leaching during burial. SRL 165 + 29.8? TDS was the most
durable glass among the three compositions under investigation.
Compared with SRL 131, SRL 165 glass frit introduced 7.1 wt? more
SiC>2 and 5.7 wt? less B202 into the final glass composition (Table
4-1). After 2 years of burial, the SRL 165 + 29.8? TDS glass/glass
interface was leached 0.5 pm deep, as shown in Fig. 6-3. which is
only 1/6 of the leach depth for SRL 131 + 29.8? TDS glass. As
revealed by SIMS analysis (Fig. 6-3), the leach depths of two SRL 131
glasses are 3.6 pm (with 29.8? TDS waste) and 1.85 pm (with 35? TDS),
respectively. These data show that SRL 165 frit is much better than
* Hydrated amorphous alumino-silicate.

TIME (MONTH)
Fig. 6-3. Penetration depth as a function of leaching time for the SRL glasses either buried in
contact with glass, stainless steel, granite or bentonite in Stripa mine, or leached in
Stripa ground water with SA/V =0.1 or 1.0 cm-1 in laboratory.

163
SRL 131 and an increase in waste loading from 29.8 wt$ to 35 wt$
decreased the leach depth of SRL 131 glass by 2X.
There were no obvious effects on SRL glass leaching due to the
presence of Cu, stainless steel or Ti. McVay and Buckwalter [54]
have reported that a synergistic effect occurs between ductile iron
and PNL 76-68 borosilicate glass: iron enhances glass dissolution
and glass enhances iron corrosion. This is due to the formation of
an iron silicate precipitate. The precipitate removes elements such
as Si from solution and therefore inhibits the saturation effects
which normally cause decreases in elemental removal rates. Thus, use
of any form of iron in the repository might be a cause for concern.
However, stainless steel used in these experiments exhibited
excellent resistance to the laboratory and burial environments. The
stainless steel (304L) used in the Stripa burial experiment belongs
to group III austenitic type which possesses better corrosion
resistance than the straight chromium (groups I and II) steels
[82]. The stainless steel did not appear to be degraded in either
the burial or laboratory experiments.
Regarding the glass/granite interface, the leach rates are
different from spot to spot. As shown in Fig. 6-3, large variations
are observed for SRL 165 + 29.8$ TDS glass surface in contact with
granite during a 2-year burial. Some areas of this glass that were
in contact with granite had larger leach depths than those on the
glass buried in contact with the bentonite. It should be pointed out
that the most heterogeneous attack occurred on the samples buried in
contact with granite for 2 years.

The presence of bentonite accelerated the attack on SRL glasses
during the first year of burial at 90°C [61,83]- However, 2-year
experimental results did not show any marked difference in the extent
of leaching between the glass exposed to bentonite and that exposed
to other types of materials (Fig. 6-3). The fact that bentonite did
not significantly influence the long-term SRL glass alteration is
probably due to saturation of the ion exchange capacity of the
bentonite. As shown in Fig. 6-4, the Li and Na depletion depths are
both larger than that of boron, obviously through an ion-exchange
process in the presence of bentonite. This may be compared with the
profiles at the glass/glass interface in Fig. 5-23 where the
"shoulder" for B practically coincides with those for Na and Li.
However, when the ion exchange sites in the bentonite structure are
substituted (e.g., A1 ions in the octahedral sites in bentonite are
substituted by Li from the glass), the effect of bentonite on alkali
extraction is reduced. The increase in the Si concentration in the
leached glass surface may be also due to the presence of bentonite,
since SiC>2 may be transferred from the bentonite to the glass surface
through a dissolution-precipitation process.
Heterogeneities can affect the corrosion behavior of alkali
borosilicate simulated nuclear waste glass during burial. This work
has shown that the enhanced aqueous attack of the heterogeneous
nuclear waste glass can occur at the phase boundaries (see
Fig. 5-26). The preferential phase boundary attack is likely to be a
result of stresses surrounding the embedded crystals and/or a

Gramatoms Remaining Based on IOO
Gramatoms of Unleached Glass
165
DEPTH (pm)
Fig. 6-4. SIMS compositional profiles of SRL 165 + 29.8% TDS
glass/bentonite interface after 24-month, 90°C burial in
Stripa (compare the corresponding glass/glass profiles in
Fig. 5-23).

166
compositional gradient between the crystalline phase and glassy phase
created during cooling.
For homogeneous glasses, there is only one contribution to
corrosion, i.e., the glassy matrix. In the case of heterogeneous
glasses, however, there are three contributions to corrosion: (1)
the glass matrix, (2) the crystalline phase and (3) the interfacial
regions between the two phases. The total extent of corrosion will
be the sum of these contributions. In the order of decreasing
durability are the crystalline phase, the glassy phase and the
interfacial regions. As discussed by McCracken [84], the relative
contributions to corrosion depend on the quantities of
heterogeneities. When small volume fractions of crystallites are
present, the improvement of chemical durability provided by the
crystalline phase is less than the reduction in the chemical
durability due to the creation of the interfacial regions. With
large volume fraction of crystals and large grain sizes, the crystal
phase will dominate the total extent of corrosion so that an
improvement of chemical durability can be expected [84].
Figure 6-5 summarizes the various modes of corrosion in the
heterogeneous alkali borosilicate simulated nuclear waste glass
observed in the burial tests. They include (1) leaching of the glass
matrix, (2) enhanced attack of the glass-crystal interface, (3)
pitting of the polycrystalline phase at grain boundaries, (4) surface
films enriched in the less soluble multivalent species and (5)
crystallite stripping due to preferential attack of the glass crystal
interface. Each of these modes of corrosion should be considered

167
(c) (e)
y
(b)
RWj
(d)
Fig. 6-5. Five modes of corrosion in partially devitrified alkali
borosilicate simulated nuclear waste glass: (a) leaching
of the glass matrix; (b) enhanced attack of the glass-
crystal interface; (c) pitting of the polycrystalline
phase at grain boundaries; (d) surface films enriched in
the less soluble multivalent species; and (e) crystallite
stripping.

168
when describing the effects of heterogeneities on glass leaching.
The dominant mechanism will depend on several factors, including bulk
composition and the volume fraction of heterogeneities.
A Model of Alkali Borosilicate Glass Leaching
The concept of using glass as a host for radioactive waste is
based upon the radionuclides entering into and becoming part of the
random three-dimensional glass network. The structural network of
the glass is provided primarily by [SiO^,]^-, [BO^]^ and [BO^]^-
polyhedra. Neighboring polyhedra are bonded together by sharing
strong ionic-covalent bridging oxygen bonds. Other multivalent
+ O + P
species such as Fe ’ rare earths or actinides are also generally
bonded within the network by bridging oxygen bonds. Low valence
ions, such as Na+, Cs+, Sr+2, etc., are incorporated into the network
by sharing various nonbridging oxygen bonds, the bonding also
depending upon size of the ions. This difference in type of bonding
in the glass network is responsible for the complex leach behavior of
nuclear waste glasses [15].
Three distinct features of glass leaching must be examined,
understood and integrated into a coherent dynamic picture. The first
of these is the mechanism of attack of the aqueous leachant on
glass. The primitive processes that must be considered and
elaborated include corrosion of the glass matrix and diffusion-
controlled migration of mobile species through the matrix to the
glass-leachant interface. The second is the control of the leaching
process by the solubilities of the various glass components in the

169
leachant. Lastly, the impact of an altered surface layer must be
evaluated. Of course, these three features are not to be regarded as
independent. One must examine the interplay of all the three as they
interact synergistically, to determine the behavior of glass
leaching.
Table 6-3 lists the dissociation energy, coordination number and
bond strenth of most oxides in alkali borosilicate nuclear waste
glasses. Based upon their bond strengths, which were calculated by
Sun [85], oxides can be divided into three categories: (1) network
formers, such as SÍO2 and 820^, whose bond strengths are above 80
kcal/gram*atom constitute the backbone of glass; (2) network
modifiers, such as Na2Ü and L^O, whose bond strengths are below 60
kcal/g.atom occupy random positions distributed through the structure
and are located to provide local charge neutrality; and (3)
intermediates, such as A^O^, whose bond strengths are between 60 and
80 kcal/g.atom may contribute in part to the network structure. The
values of the bond strength or dissociation energy indicate the
hierarchy of energy required to break various bonds in the glass
network.
In reference to the three major cations in alkali borosilicate
nuclear waste glasses, Si^+, B^+ and Na+, note that since Na ions are
modifiers they are mobile in the glass structure and can be released
through an ion-exchange process in an aqueous solution, as described
by the following equation:
[-Si-ONa], . + Ho0, . = [-Si-OH], , + Na, . + OH. . (6-1)
I (g) 2 (aq) , J(g) (aq) (aq)

170
Table 6-3. Coordination Number and Bond Strength of Most Oxides in
Alkali Borosilicate Nuclear Waste Glasses.
M in
M0X
Valence
Dissociation
Energy per M0X
(kcal/g*atom)
Coordi¬
nation
Number
Single-Bond
Strength
(kcal/g «atom)
Glass formers
B
3
356
4
89
B
3
356
3
119
Si
4
424
4
106
A1
3
402-317
4
101-79
Zr
4
485
6
81
Intermediates
A1
3
317-402
6
53-67
Zr
4
485
8
61
Ti
4
435
6
73
Modifiers
Li
1
144
4
36
Na
1
120
6
20
K
1
115
9
13
Cs
1
11 4
1 2
10
Mg
2
222
6
37
Ca
2
257
8
32
Sr
2
256
8
32
La
3
406
7
58
Adapted from [85].'

171
As a result, the solution will have a concomitant increase in pH. An
introduction of a modifier oxide such as Na20 reduces glass melting
temperature, but increases the number of nonbridging oxygen bonds in
the glass structure and reduces chemical durability.
Removal of B should involve breaking 3 or 4 bridging oxygen
3- 5-
bonds associated with [BO^] or [BO^] units. The following
equations represent breaking one bridging oxygen bond associated with
the B structural units.
II I I
[-B(Na)-0-B-], , + Ho0, , * [-B(Na)-OH], . + [-B-0H], . (6-2)
| (g) 2 (aq) | (g) (g)
II II
C-Si-O-B(Na)-]. . + HO, . - [-Si-OH], . + [-B(Na)-0H], . (6-3)
j| (g) 2 (aq) | (g) | (g)
Since the dissociation energy of Si02 (424 kcal/g*atom) is
larger than that of B^pO^ (356 kcal/g«atom), release of Si is
energetically less favorable. Breaking of Si-0 bonds can be
represented by
I l I
[-Sl-0-S1-](g) . H20(aq) < 2E-Si-0H](g) (6-4)
This is one of the reasons why all the data shown in Chapter V
indicate preferential leaching of boron.
The second reason for preferential leaching of boron is the
large difference in stability between boria and silica in aqueous

172
solutions. As pointed out by Paul and Cooke [86] and shown in Fig.
6-6, the activity of is several orders of magnitude larger than
that of vitreous Si02. Thus, it is unlikely that reaches its
solubility limit. Although it has been reported that several boron-
containing silicate phases were identified on the altered glass
surface, these phases were shown to be artifacts of the hydrothermal
experimental method in which the leachates were allowed to evaporate
in contact with the altered glass [87]. On the other hand, the
silicic acid dissolved from the glass may reach saturation and
precipitate to form Si02, as described in the equation below.
HHSi0l|(aq) * S1VH2°(g) * H2° * S1°2(g) * 2H2° (6'5)
It should be mentioned that the SÍO2 formed through the condensation
reaction (equation (6-5)) is different from the Si02 in the original
glass, but it is still amorphous for glasses leached at temperatures
not higher than 90°C and at 1 atmosphere as revealed by x-ray
diffraction. The existence of strong ionic-covalent bridging oxygen
bonds, especially Si-O-Si bonds, is most likely responsible for the
low leachability of many nuclear waste glasses over a pH range from
4.5-9.5.
Selective dissolution of glass network formers plays an
important role in glass leaching. Preferential dissolution of
boron-containing units and preservation of the Si02 network result in
an altered glass surface rich in SiC>2* Such a layer may serve as a

173
pH
Fig. 6-6. Stability of B2C>3 and Si02 in aqueous solution at 25°C as
a function of pH (adapted from [86]).
l

174
barrier to reduce the release of other species including from
the glass. Release of does not necessarily increase OH
concentration as does that of Na20 (compare equations 6-1 and 6-2).
The solubility of silica is low in a solution of low pH.
All of the alkali ions are not released through an ion exchange
process. When the exchange site is Si-0-Na+, the thermodynamics of
the exchange reaction are favorable due to the high strength of the
OH bond which forms. However, Na contained in [B(Na)Oii]Z<- sites will
not take part in ion exchange process since the negative charge on
this site is associated with the tetrahedron rather than a particular
oxygen, and unpolarizable cations such as H+ cannot effectively
compensate for this charge as well as can the Na+ [88]. The
thermodynamics do not favor H+-Na+ exchange at these sites. Thus,
extraction of Na from [BiNaJO^]^ sites can only occur through the
release of the whole units.
Some metal ions originally contained in the ground water or
released from the glass may absorb onto the glass surface and be
concentrated within the altered layer. They change the solubility of
silica by forming less soluble metal silicates. A good example of
such metal ions is Al^+. It was found that, in most cases, A1 is one
of the least mobile species in the glass (see the A1 profile in Fig.
4-17). Aluminum ions can absorb onto the hydrated silica surface to
form anionic aluminosilicate. The absorbed sites can be either on
the outer surface or within the altered layer. This results in
drastic reduction of the dissolution rate as well as the equilibrium
solubility of the altered glass [77].

175
Breaking of the structural units in alkali borosilicate nuclear
waste glass results in dissociation of all other species dissolved in
it and linked together with less strong bonds. Then, based on the
solubility limits of metal ions in aqueous solution, multiple surface
layers of less soluble elements form. The onset of surface
precipitation depends on the time required for various species to
reach saturation in solution with respect to the surface complex.
Saturation of species "i" will be a function of the initial pH,
amount of alkali in the glass and rate of alkali release,
temperature, initial concentration of species "i" in the solution,
SA/V which influences solution concentration and flow rate which also
affects solution concentration.
In summary the proposed model of alkali borosilicate glass
leaching is based on the chemical bonding of various M-0 connections
and the stability of their individual oxides in aqueous solution.
Figure 6-7 (a), a schematic of this model, shows a cross-section
of the altered glass surface and typical compositional profiles
observed for Si, B and alkalis. It is postulated that most Si02 in
the leached layer should be embodied within an unbroken three-
dimensional network. The two major network former ions, Si and B,
are released at different rates, and preferential leaching of boron
is obvious. The [SiO^]^ units preserved in the glass surface can
hold other species contained therein, keeping them from release into
solution. At the glass-water interface, boron depletion is almost
complete. Leaching of alkali borosilicate glass is characterized by
this selective dissolution of network formers. Preferential

176
Surf.
M
Network
Dissolution
(a)
Fig. 6-7. Schematics showing (a) the altered alkali borosilicate
glass surface and the compositional profiles after leaching
based on the model proposed in this dissertation and (b)
the altered glass surface based on Grambow's model.

177
Complete network dissolution
with preciptation
(b)
Fig. 6-7—continued.

178
leaching of boron and preservation of the SiC>2-rich skeleton result
in a honeycomb structure (in micron range) where glass density
drops. The least soluble species entering solution after breaking
down the Si- and B-containing units will form precipitation layers
either on the outer surface or within the leached layer. The larger
depletion depth of alkalis compared with that of boron is due to an
ion-exchange reaction. However, high concentration of alkalis in
solution may suppress the ion-exchange reactions.
As mentioned in Chapter II, Grambow proposed a model of glass
leaching based on the role of metal ion solubilities. Grambow's
model became the theoretical basis for understanding the surface film
formation at the nuclear waste glass surface. However, Grambow did
not address selective network dissolution for the borosilicate
glass. In his model all the network former ions, Si and B are
released at the same rate. Thus, the ratio of the normalized mass
loss of B to that of Si is equal to 1 (Fig. 2-4). Current data
indicate that this is not always the case. In contrast, preferential
network dissolution of B is the common feature observed in static and
flow leach tests. If Grambow's idea was correct, the curves shown in
Fig. 5~35 should be overlapped on each other. In contradiction to
the Grambow's model, the normalized leach rate of B is larger than
that of Si. However, the model given in this dissertation can
explain the nature of selective dissolution of the network formers
quite well.
Secondly, Grambow's model assumes that all the constituents in
the glass must totally dissolve into solution. This means that, in

179
the leaching history, the formation of the altered layer required
that the glass must dissolve congruently at least once. When
congruent dissolution raises solution concentration of certain
elements to the level at which new solid phase form, these phases
will regulate the solution concentration. Grambow's model is
schematically shown in Fig. 6-7 (b). The surface films form due to
precipitation only after complete network dissolution.
Figure 6-8 shows the density at the altered layer as a function
of depth for the three SRL glasses after 2-year burial in Stripa at
90°C (glass/glass interface). The decrease in density is basically
due to leaching of Na, Li and B. The structure of the altered layer
appears more open than that of the bulk glass and contains
micropores. The density drop is, therefore, directly related to the
release of Na2Ü, L^O and B2O2 in the glass. The number and size of
these micropores may change with depth and also depend on the
composition of glass and solution. The durable glasses produce a
thin coherent dense surface layer which can protect the glass more
effectively from further leaching. Precipitates of less soluble
species on the glass surface regulates the density. This behavior is
shown in Fig. 6-8 by the local increase in the density at - 0.3 pm
depth of the SRL 131 + 29.8% TDS glass specimen.
Effect of Glass Composition
Figure 6-9 is the compositional ternary diagram showing six
alkali borosilicate nuclear waste glass compositions. Three regions
can be identified. These range from low leaching systems

DENSITY INDEX
1.2
SRL 165+ 29.8% TDS
0.8
0.6 -
0.4
0.2
- 2.5
- 2.0
.0 2.0
DEPTH (/xm)
Fig. 6-8. The density index curve for three SRL glasses after 2-year burial in Stripa at 90°C.
BULK DENSITY (g/cm3 )

181
Si 02
1 - SRL 131 + 29. 8 % TDS
2 - SRL 131 •*- 35% TDS
3 - SRL 165 + 29.8% TDS
4 - ABS 39
5 - ABS 41
6 - ABS I 18
Fig. 6-9. Compositional ternary diagram showing the direction of
increasing boron depletion depth. R20 represents alkali
metal oxide, MegO^ represents and FepO^ and WP
stands for waste products.

182
(-0.4 pm/year) to intermediate leaching (1-1.2 pm/year) to high
leaching systems (2.5-3 pm/year), based on the boron depletion depths
of the glass/glass interface for the 1-year burial samples. The
leaching data were correlated with the (Si02 + A^O^/O^0 + B^O
wt% ratio, where R20 stands for alkali oxides. Since preferential
leaching of R20 and B^jO^ are observed with all glass compositions and
all glass/repository materials interfaces as addressed in the model,
a sum of R20 + B^^ is used as the denominator in the ratio. Figure
6-10 is a plot showing the boron depletion depth as a function of
(Si02 + A1202)/(R20 + B20g) wt ratio for these glasses. It was found
that leach resistance increased as this ratio increased. SRL 165 +
29.8? TDS exhibited the smallest leach depth of boron among the six
waste glasses. This is not surprising since this glass composition
is nearest to the Si02 - (Me-jO^ + WP) edge where Me^jO^ stands for
Al202 and Fe-jO^, and WP, the waste products, and contains the least
amount of alkali and boron with the highest (Si02 + A1202)/(R20 +
B2O2) ratio of 2.44 (see Fig. 6-10).
Glass ABS 41 is a good example to show how well the proposed
model explains the leach data. Although this glass contains nearly
the same amount of Si02 and A^Og (54.5%) as SRL 165 + 29.8% TDS
(54.6%) and less Na20 + Li20, the doubled content of B^^ (15.90% in
ABS 41 vs 7.1% in SRL 165 + 29.8% TDS) is primarily responsible for
the doubled leach rate of ABS 41 as compared to SRL 165 + 29.8%
TDS. Based on Grambow's model, since these two glasses contain

183
Fig. 6-10. The boron depletion depth as a function of (SÍO2 +
A^O^/U^O + B2O2) wt ratio in glasses.

184
nearly the same amount of SÍO2 + Al20g, their leach rates should be
the same.
Among the six glass compositions investigated, the least durable
glass was ABS 39 with the lowest (Si02 + Al-pO^/tR^ + BgO^) ratio of
1.57 and the highest amount of B^^ (19.12%). Also, this glass
contains <1% Cs20 and no Li20. Therefore, a mixed alkali effect
[78,80], if present, was not significant. The second least durable
was SRL 131 + 29.8Í TDS with a (Si02 + Al-pO^/Cf^0 + B20^) ratio of
1.52 (Fig. 6-10). This SRL glass contains the highest amount of
(Na20 + Li20 + Cs20) and second least amount of (Si02 + A120^). The
other two compositions, SRL 131 + 35% TDS and ABS 118, are
categorized in the intermediate leaching systems together with ABS
41. Glasses ABS 118 and ABS 41 have the same (Si02 + A1203)/(R20 +
B-pO^) ratio (1.87). However, ABS 41 contains 4.1 % more (Si02 +
A^O^) and 2.2% more alkalis and B-pO^ as compared to ABS 118. As
shown in the compositional ternary diagram (Fig. 6-9), ABS 41 is
closer to the Si02 - (R-,0 + B20g) edge and slightly closer to the R20
+ corner than ABS 118. An improved durability of SRL 131 + 35%
TDS was due to the higher waste loading, although this glass has the
second lowest (Si02 + A12Ü2)/(R20 + B-pO^) ratio (1.69). One of the
major factors responsible for this improved durability is that the
simulated SRP nuclear wastes consist primarily of Fe^^i MnO and
Al-pO^ which have low elemental leachability. This is because these
species decrease the solubility of silica in the solution by forming
a coherent surface layer of less soluble metal silicates on the glass

185
surface [77]. Thus, in a well preserved Si02 network, preferential
leaching of boron will be more difficult. The second factor is due
to the higher waste loading which lowers the absolute concentration
of and increases the average distance between two nearest boron-
containing units in the glass structure.
Influence of Repository Variables
Ground Water Chemistry
Laboratory experiments (Fig. 5~30) have shown that the extent of
surface deterioration is less in ground water than in deionized water
and that water which already contains large concentrations of glass
species is less aggressive than regular ground water or deionized
water. Results also show that the major factor contributing to the
reduction in leaching was the presence of Si in solution (Fig. 5-30).
As shown in equation (6-11), an increase in Si concentration will
suppress the reaction in the right direction. As a result, the rate
of Si dissolution will be lower in Stripa water containing 13 ppm of
Si than in deionized water. This effect is illustrated schematically
in Fig. 6-11 where the concentration vs time curve is shown. The
leach rate of silica from the glass can be represented by the slope
of the curve. The greatest slope, and hence the largest leach rate,
occurs when the concentration of silica in solution is equal to
zero. In a static leach test, the leach products accumulate in
solution and eventually approach saturation. The leach rate of
silica decreases as its solution concentration increases. When the

CONCENTRATION
186
CONTACT TIME
Fig. 6-11. Schematic illustrating
concentration, contact
[36]).
B.
C.
the relationship between
time, and leach rate (adapted from

187
solution becomes saturated with respect to silica, its leach rate, as
measured by its change in solution concentration, will be zero.
Based on this explanation, it can be seen from Fig. 6-11 that the
leach rate of silica in deionized water, where its initial
concentration is zero (C^). will be greater than the leach rate of
the same element in a leachant with an initial concentration equal to
C2 or C-| . Of course, each element has a different saturation
concentration, but since silica is the major structural component of
the glass a reduction in its dissolution will result in concomitant
reduction in dissolution of the other glass constituents.
It should be pointed out that the rate of B release depends to a
large extent on the rate of SÍO2 dissolution. This is due to the
fact that in the glass structure, [SiO4]14_, [BO^]-’- and [BO^]^- are
connected through bridging oxygen bonds into an integral three-
dimensional network. Dissolution of more SÍO2 structural units
results in more release of [BO^]'’ and [BO^]^ units, and vice versa
is also true.
Secondly, a pH buffering effect of Stripa ground water has a
strong effect on waste glass leaching. Stripa ground water contains
weak bases and their corresponding salts such as Ca(0H)2 and CaC^»
and FeCOH)^ and FeCl^ (see Table 4-7). The OH- ions produced as a
result of alkali release from the glass can react with Ca and Fe to
form corresponding bases. Therefore, during the burial tests, the
ground water pH was found to increase by only 1 unit. The laboratory
simulations using granite rock cups and Stripa water in both static
and flow conditions (at <0.3 mL/h) also show that the ground water pH

188
varied between 7.3 and 8.4 (Fig. 5-32). In contrast, the pH of
leachant in the MCC-1 static test using deionized water reaches 9.54
after 28 days and 9.41 after 1 year [27].
Stripa ground water contains some monovalent cations such as Na
and K. The existence of these monovalent cations in solution can
reduce the solution pH by ion-exchange with the glass surface silanol
protons [89]. The- data of Shade et al. [90] show that hydrolysis of
glass matrix was generally slower in brines than in deionized water,
due to a lower solubility of silica at lower pH in brines. Thus, in
Stripa ground water, SÍO2 dissolution rate is expected to be low
except in cases where the glass is highly rich in alkalis.
Effects of Repository Materials
Early results of the Stripa burial experiments [61 ,83] have
shown that the presence of bentonite resulted in an accelerated
attack on glasses within 1-year burial at 90°C. Recent data
collected from this experiment indicate that bentonite has less
effect on long-term glass leaching. This trend is best shown in Fig.
6-1 (a). When it was in contact with bentonite during burial, ABS 39
leached at a high rate during the first 3 months, then slowed
dramatically and kept the same low leach rate through the 31st month.
There are several reasons which could account for more severe
attack on the glass surface exposed to bentonite during an earlier
period of burial:
(1) Experimental data show that the pH of bentonite-containing
ground water is 0.5-1.4 units higher than without bentonite. Thus,

189
higher steady state concentrations of SÍO2 would be expected in the
presence of bentonite compared to water alone. The early surface
roughening as shown by low intensity of the FT-IRRS spectra (Fig.
5-7) is indicative of an accelerated dissolution of Si02 due to the
quick rise of pH in the bentonite containing ground water. The
characteristics of selective network dissolution is less apparent.
(2) Enhanced ion exchange reactions existed between glass
surfaces and bentonite. Bentonite provides a solution saturated in
soluble Ca-Mg, Fe, Al, etc. and a high base exchange capacity for Na
[19]. As shown in Fig. 6-4, Na, Li and Cs leached up to 1-1.3 ym
deep into the surface with the Ca concentration build-up to nearly
the same depth. The depletion of the alkalis is in contrast with the
case of glass/glass interface (Fig. 5-23) considerably deeper than
that of B.
The time-dependency of the leach depth based on boron extraction
for ABS 118 glass/Pb, glass/Cu, glass/Ti, glass/glass and
glass/granite interfaces after 90°C Stripa burial is compared in Fig.
6-12. Generally, very low leachability (_<3-48 ym/yr) was found on
all these glass surfaces after 12-month, 90°C Stripa burial. It is
seen that glass which was in contact with Pb exhibits the smallest
leach rate among the five glass interfaces. Thus after -300 years of
the thermal storage period, glass which is in contact with Ti will
leach to -1,000 ym depth while glass in contact with Pb will leach
only 40 ym. As shown in Table 6-4, the glass/Ti interface leached by
a factor of 7 faster, while both glass/Pb and glass/Cu interfaces
corroded by a factor of 4 slower than the glass/glass between 7-12

BORON DEPLETION DEPTH (jjm)
1 90
TIME (MONTHS)
Fig. 6-12. The boron depletion depth as a function of burial time
for ABS 118 glass/glass, glass/granite, glass/Pb,
glass/Cu and glass/Ti interfaces after 90°C Stripa
burial.

191
Table 6-4.
90°C ABS Glass
Leach Rates During
7-12 Months
Period.
Interface
Glass/Glass
Glass/Pb
Glass/Cu
Glass/Ti
Leach rate
(ym/yr)
0.48
0.12
0.12
3.48

192
months of burial. The leach depths after 12-month 90°C Stripa burial
are (in increasing order) glass/Pb < glass/granite < glass/Cu <
glass/glass < glass/Ti.
One of the factors which may be responsible for the different
leach depths with different glass/overpack metal interfaces is the
difference in the tightness of the contact between the glass and
various metal surfaces. Since the compacted bentonite which was used
as the backfill material in the Stripa burial experiment absorbed
ground water and swelled, high pressure was produced within the
boreholes. The flexibility of some overpack metals such as Pb
permits the metals to press tightly against the contacted glass
surface under such high pressure and thus the ground water in the
gaps between glass surfaces and metals was severely limited. Another
possible reason why the presence of Pb may reduce glass leaching is
that Pb dissolved from the metal coupon can adsorb onto the glass
surface and form less-soluble lead silicate [77]. Thus, the amount
of SiC>2 which needs to be dissolved from the glass decreases.
As shown by the SIMS analysis (Fig. 5—16), there was no Si02
network dissolution observed at the three glass interfaces. An
apparent decrease in the Si concentration was found only at the
<0.2 pm leached layer of the glass/Pb. This may possibly be an
artifact of the SIMS data processing due to smudges of metallic Pb on
the specimen surface. Considerable amounts of Na and B still remain
unleached at this <0.2 pm surface region indicating very little
surface attack.

193
Effect of Temperature
Temperature influenced glass leaching significantly during
burial. As regards the low temperature (8-10°C) leaching
characteristics, the B depletion curves based on SIMS analysis (Fig.
6-3) indicate that the depth of the SRL 131 + 29.8% TDS glass/glass
interface is approximately one order of magnitude less at 8-10°C than
it is at 90°C and that the qualitative differences in B depletion
depths between the three SRL glasses are very small. The SIMS data
(Fig. 6-13) also shows little evidence of ion exchange at ambient
temperature as revealed by the H profile. Thus, it appears that at
8-10°C, network dissolution is the predominant leach mechanism for
this nuclear waste glass composition. This argument is in
contradiction to that proposed by Zagar and Schillmoeller [91] which
states that at temperatures below 30°C ion exchange is predominant.
Comparison of Field and Laboratory Test Results
In order to approximate the Stripa burial conditions laboratory
tests were conducted using Stripa ground water, granite rock cups and
higher SA/V ratios. Results show that the Stripa repository
conditions could be approximated in the laboratory using SA/V £1 .0
cm 1 and static or low flow (£0.1 mL/h) conditions. An increase in
the flow rates from 0.1 to 0.3 mL/h resulted in a 6X increase in the
leach rates based on the mass loss data shown in Fig. 5~3^. The SA/V
ratio in the Stripa burial test at glass/glass interfaces was
probably >5 cm-1 depending upon the ground water accessibility and
the extent of bentonite intrusion.

Gramatoms Remaining Based on
IOO Gramatoms of Unleached Glass
1 94
SIMS compositional profiles of SRL 165 + 29.8% TDS
glass/glass interface after 8-10°C Stripa burial for 2
years.
Fig. 6-13.

195
The SIMS depth profiles shown in Fig. 5-23 (b) for SRL 165 +
29.9% TDS glass buried for 2 years in Stripa at 90°C, and Figs. 5-31
and 5-37 for the similar glass leached statically in the laboratory
in ground water with SA/V = 1.0 cm 1, suggest that the glass leached
through similar corrosion mechanisms in the field as in the
laboratory. These profiles indicate that depletion of alkalis and
boron, and to a more limited extent silica, had occurred in both the
laboratory- and field-leached glasses. At the outer surface, the
concentrations of Na, Li, B and Si were significantly depleted with
respect to the unleached glass. Within the inner region, the alkalis
and boron had been leached to a greater extent than silica, and B
appears to have been leached to a lesser extent than the alkalis,
such as Li. In addition, the curves for the same elements in these
two cases assume nearly the same shapes, suggesting that the leach
mechanisms under the field and laboratory conditions are most likely
the same.
Regarding the leaching kinetics, the granite rock cup tests
simulated the Stripa granite repository much better than the
laboratory test in which granite cup was not used. As indicated by
the boron depletion, the leach depths after 1-, 3- and 6-month
leaching under the granite rock cup test conditions are 0.16, 0.24
and 0.45 um (Fig. 5-37). The boron depletion depth of the 1-month
laboratory sample leached in absence of granite is 0.46 pm. This
indicates that glass leaches at a lower rate in an environment
containing granite. It is expected that if a higher SA/V ratio is

196
_ I
used (=5 cm ) in the granite rock cup test, glass leach rate will be
closer to that observed in Stripa burial tests. Data from both
field- and laboratory-corroded glass samples confirm very well the
validity of the proposed model of glass leaching.
The weight loss of a SRL 165 burial glass slice calculated from
2 2
Fig. 6-8 is approximately 0.31 g/m , as compared with 0.25 g/m
(based on sample weight) for a 28-day laboratory-leached sample using
ground water with SA/V = 1.0 cm 1 at the same temperature.

CHAPTER VII
SUMMARY
Burial experiments with three SRL and three ABS simulated
nuclear waste glasses were conducted to evaluate the resistance of
these glasses to ground water attack under repository-like
conditions. Glass samples were buried in the boreholes at a level of
-35O meters below the surface in the Stripa granite at either ambient
mine temperature (8-10°C) or 90°C. Included in the same boreholes
were potential waste package components. Two sample configurations,
pineapple slices and minicans, were used. Glasses were also leached
in the laboratory using the Stripa ground water and granite rock cups
to simulate the Stripa repository.
The leached surfaces were characterized using SEM-EDS, FT-IRRS,
SIMS, RBS and optical microscopy in combination. Solution analytical
techniques were also used wherever possible.
The leaching behavior of six nuclear waste glasses including
three American defense HLW glasses and three Swedish glasses
containing commercial HLW in realistic repository conditions was
evaluated. A leaching model for the alkali borosilicate nuclear
waste glasses was proposed to elucidate the mechanisms of
borosilicate glass leaching. The model is based on the
considerations of the bond strengths for different oxides contained
in the glass structure and stability of these oxides in aqueous
197

198
solution, which could explain the observed preferential leaching of
boron and preservation of the Si02 honeycomb network. This model is
different from Grambow's model which assumes complete dissolution of
all glass constituents followed by precipitation based on the
solubility limits of various oxides in the glass [39].
A significant compositional effect on glass leaching was
observed with the six simulated nuclear waste glasses under burial
conditions. The leach rate expressed by the annual boron depletion
depth was inversely correlated with (Si02 + A^O^/CF^O + E^O^) wt
ratio in these glasses which can be considered in line with either
model. Glass SRL 165 + 29.8? TDS was the most durable composition
among the six. An increase in waste loading of SRL 131 from 29.8% to
35? decreased its teachability by 2X. These results agree with the
earlier work on the compositional dependence of nuclear waste glass
leaching [92,93].
Accelerated attack during the first year in the presence of
bentonite appears to be a transient effect. The time when this
alleviation of the bentonite effect occurs is dependent on glass
composition and may be disturbed by the local effects such as
variation in the effective SA/V ratio.
Little difference in leach depths was observed between
glass/glass, glass/stainless steel and glass/Cu interfaces. The
glass surfaces which were in contact with granite showed smaller or
larger leach depths than glass/glass interface, and heterogeneous
attack was usually found on the glass/granite interface. The rock
cup tests show that the presence of granite resulted in a reduction

199
in leach rates. The slowest leach rate was found at the glass
surface which was in contact with Pb. Pb provided an intimate
contact between the glass and Pb surface after bentonite swelling,
resulting in a high SA/V environment and reduced leaching.
Results of the Stripa burial tests show that glass specimens
containing crystallites exhibit preferential attack of the interface
between crystalline and glassy phases. The crystalline phase,
identified as spinel solid solution, exhibits better chemical
resistance than the glassy phase. These results are consistent with
the laboratory leach results [9^-98]. The degradation of the leach
rates of the partially devitrified glasses is due to preferred attack
at the glass-crystal phase boundaries.
There is no obvious effect of sample configuration (minicams vs
pineapple slices) on glass leaching since the stainless steel rings
are chemically inert.
Temperature significantly influenced glass leaching during
burial. Based on SIMS depth profiling, glass leaching was 3X to 10X
as fast at 90°C as at 8-10°C, depending upon the glass
compositions. It appears that, at low temperatures, network
dissolution is the predominant leach mechanism.
Short-term laboratory static tests using ground water and high
SA/V ratios (_>1 .0 cm 1) can approximate the Stripa burial environ¬
ment. The SIMS in-depth profiles of glass surfaces after Stripa
burial and laboratory simulation tests reveal that glass leached by
the similar mechanisms, but more slowly under burial conditions than

200
under the laboratory-controlled environment. This is attributed to
the difference in the SA/V, the burial glass samples being leached
with SA/V >5 cm-1. Surface composition and morphology are nearly
identical for glasses leached under laboratory and burial conditions.
Using the slopes in the boron depletion depth vs burial time
curves, the leach depths of the glasses can be extrapolated up to 300
years of storage (the thermal period) from the 90°C SIMS data, and to
10^ years of storage from both 90°C and 8-10°C data. Tables 7-1 and
7~2 list the results of calculations. It is shown that, except for
the ABS 39 glass/glass and glass/granite interfaces and ABS 41
glass/bentonite interface, the boron depletion depths are less than
1,000 pm after 300 years of storage assuming these glasses are
exposed right after the disposal sites are closed and the stainless
steel canisters are assumed to be breached right away. The estimated
boron depletion depths of the three SRL nuclear waste glasses, which
c;
are in contact with glass of the same composition during 10 years of
burial would be less than 10 mm (Table 7-2).
All these results show that Stripa burials combined with
laboratory simulations are unique experimental designs which have
provided useful information regarding leach performances of nuclear
waste glasses. This work has served as a model on which design and
development of the MITT tests are based. Over 1,000 samples from 8
countries are involved in WIPP tests, which is thought to be the
"second generation" of the Stripa tests.

Table 7-1. Estimated Boron Depletion Depths (pm) after 300 Years of the Thermal Period of Storage for
the Six Nuclear Waste Glasses. It is assumed that glasses will be in contact with ground
water right after the repository is closed and leach at constant 90°C. Therefore, such
kind of estimation only represents the upper limits of glass leaching.
Interfaces
Glass/ Glass/ Glass/
Glass Composition Glass/Glass Granite Bentonite Stainless Steel Glass/Pb Glass/Cu Glass/Ti
SRL 165
+ 29.8% TDS
10
<20
<20
SRL 131
+ 35% TDS
200
-
-
SRL 131
+ 29.8$ TDS
300
-
-
ABS 39
2,700
5,700
400
ABS 41
10
70
3,600
ABS 118
300
10
700
40
1 ,000
201

202
after 10D Years of Storage for the
Glass/Glass Interfaces of SRL
Simulated Nuclear Waste Glasses. It
is assumed that glasses will be in
contact with ground water right after
the repository is closed and leach at
constant 90°C. Therefore, such kind
of estimation only represents the
upper limits of glass leaching.
Glass Composition
Depth
SRL 165 + 29.8? TDS
6,300
SRL 131 + 35? TDS
8,800
SRL 131 + 29.8? TDS
9,600

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BIOGRAPHICAL SKETCH
BingFu Zhu was born June 22, 1946, in Shanghai, China. He
received his elementary education in Shanghai and was graduated from
Xianming Middle School in 1962. He entered the East China University
of Chemical Technology, Shanghai, in September 1962, and received his
diploma from the Department of Inorganic Industry in September 1968.
He was an engineer in Wanshan Cement Works, China, from
September 1968 through September 1975. Subsequently, he joined the
faculty of the Department of Silicate Technology, Wuhan Institute of
Building Materials, China, for 3 years. In September 1978, he
returned to the East China University of Chemical Technology to
pursue an M.S. in glass science and received the degree from the
Department of Inorganic Materials in February 1982. In March 1982,
he entered the University of Florida to start working for a Ph.D. in
materials science and engineering.
He has published papers on chemical strengthening of glasses and
glass corrosion. He is a member of American Ceramic Society and
National Institute of Ceramic Engineers.
He married JiSi Wang on October 14, 1981.

I certify that I have read this study and that in my opinion it
conforms to acceptable standards of scholarly presentation and is
fully adequate, in scope and quality, as a dissertation for the
degree of Doctor of Philosophy.
L£X_
David E. Clark, Chairman
Professor of Materials Science and
Engineering
I certify that I have read this study and that in my opinion it
conforms to acceptable standards of scholarly presentation and is
fully adequate, in scope and quality, as a dissertation for the
degree of Doctor of Philosophy.
Graduate Research Professor of Materials
Science and Engineering
I certify that I have read this study and that in my opinion it
conforms to acceptable standards of scholarly presentation and is
fully adequate, in scope and quality, as a dissertation for the
degree of Doctor of Philosophy.
p £
Alexander Lodding
Professor of Materials Ph
Institute of Technolog
sics, Chalmers
Sweden
I certify that I have read this study and that in my opinion it
conforms to acceptable standards of scholarly presentation and is
fully adequate, in scope and quality, as a dissertation for the
degree of Doctor of Philosophy.
Christopher D. Batich
Associate Professor of Materials Science
and Engineering

I certify that I have read this study and that in my opinion it
conforms to acceptable standards of scholarly presentation and is
fully adequate, in scope and quality, as a dissertation for the
degree of Doctor of Philosophy.
Associate Engineer of Materials Science
and Engineering
I certify that I have read this study and that in my opinion it
conforms to acceptable standards of scholarly presentation and is
fully adequate, in scope and quality, as a dissertation for the
degree of Doctor of Philosophy.
Gar B. Hoflund 1/
Professor of Chemical Engineering
This dissertation was submitted to the Graduate Faculty of the
College of Engineering and to the Graduate School and was accepted as
partial fulfillment of the requirements for the degree of Doctor of
Philosophy.
May 1987
a • /i
Dean, College of Engineering
Dean, Graduate School

UNIVERSITY OF FLORIDA
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