Nuclear waste glass leaching in a simulated granite repository

MISSING IMAGE

Material Information

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

Subjects

Subjects / Keywords:
Leaching   ( lcsh )
Radioactive waste disposal in the ground   ( lcsh )
Materials Science and Engineering thesis Ph. D
Dissertations, Academic -- Materials Science and Engineering -- UF
Genre:
bibliography   ( marcgt )
non-fiction   ( marcgt )

Notes

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

Record Information

Source Institution:
University of Florida
Rights Management:
All applicable rights reserved by the source institution and holding location.
Resource Identifier:
aleph - 000948893
oclc - 16904849
notis - AER1000
sobekcm - AA00004851_00001
System ID:
AA00004851:00001

Full Text

















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


1987

































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.
















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










Analytical Techniques..................................67
Solid State Analyses...............................67
Solution Analyses..................................85

V TEST 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

VI DISCUSSION............................................152

ABS Glasses........................................... 152
SRL Glasses........................... ............... .161
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...................188
Effect of Temperature.............................193
Comparison of Field and Laboratory Test Results........193

VII SUMMARY................................................ 197

REFERENCES.......................................................203

BIOGRAPHICAL SKETCH.............................................212















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...................................... .8

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 Gram-Atoms of Unleached ABS 118 Glass after
12-Month, 900C Burial in Stripa............................119









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 900C 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, 900C 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 900C 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 105
Years of Storage for the Glass/Glass Interfaces of SRL
Simulated Nuclear Waste Glasses............................202















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.....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 NLSi (20 g/m2)
for CaCO SrCO, Nd(OH)3, Fe(OH)3 and Zn(OH)2
in MCC-1 28-day test at 90OC in solutions of different
pH....................................................... 26

2-5 The Si leachability of a borosilicate glass immersed
in a 5-day static 230C 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 functi n of leaching time. A diffusion
coefficient of 10ut 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 (10OX)....................................... 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, 900C 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, 900C 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










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, 900C Stripa burial.....................92

5-4 SIMS depth profiles for (a) ABS 39 (Al-corrected) and
(b) ABS 41 (Si-corrected) after 31-month, 900C 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, 900C
Stripa burial..............................................97

5-6 SIMS depth profiles of boron for glass ABS 39 (a)
and glass ABS 41 (b) after 31-month, 900C 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 900C 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 900C 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, 900C burial in Stripa. The data
have been corrected using Al 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, 900C 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
900C burial in Stripa.................................... .108









5-12 Light micrographs of ABS 118 glass surfaces after
9000C 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, 900C
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, 9000C
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, 900C
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, 900C 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, 9000C 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, 900C 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, 900C 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, 900C 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 900C
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 900C......................................124











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 900C: (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 900C: (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 900C: (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 900C: (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 900C 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 cm1 and (d) saturated Stripa ground water
-1
with SA/V = 0.1 cm 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 9000C in Stripa water with SA/V =
1.0 cm ..................................................139

5-32 Solution pH vs time for Black Frit 165-Mobay glass
corroded under static and flow conditions..................141

5-33 Concentrations of Si, B, Al 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










5-34 Normalized leach rates of Li, B and Si as a function
of time under flowing (at 0.3 ml/hr) conditions for
-1
Black Frit 165-Mobay glass with SA/V = 1.0 cm1.
Also shown are the weight losses for glass samples
leached under static and flow conditions with SA/V
-1
= 1.0 cm ................................................144

5-35 EDS analysis of Black Frit 165-Mobay glass leached in
-1
the rock cup test at 900C with SA/V = 1.0 cm1 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 900C 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 900C 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, 9000C 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............................160

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
-1
cm1 in laboratory......................................162

6-4 SIMS compositional profiles of SRL 165 + 29.8% TDS
glass/bentonite interface after 24-month, 900C 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 B203 and Si02 in aqueous solution
at 250C 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 900C...........................180

6-9 Compositional ternary diagram showing the direction of
increasing boron depletion depth. R20 represents
alkali metal oxide, Me203 represents A1203 and
Fe203, and WP stands for waste products...................181

6-10 The boron depletion depth as a function of
(SiO2 + A1203)/(R20 + B203) 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 900C Stripa burial.......190

6-13 SIMS compositional profiles of SRL 165 + 29.8% TDS
glass/glass interface after 8-100C Stripa burial for
2 years.................................................. 194















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

(8-100C) or 900C. 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










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 leachability by a factor

of 2.

The leach rates of buried samples based on boron extraction at

900C ranged from 0.3-3 pIm/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 ). 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.















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





























r-i


co


ca0 L
a, c 4
p> 4 O)






0


o0 a
-1
a.)



(OM C a
xx


XE bO
aOa,















OL
> 0

a











04T 04


*-
--




a3
M


n4







Scf


CO
4-)










c,
aI-
4-3 .









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, Cs137 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

incorporation in body tissues, especially Sr90 in bones. When

decaying, they give off both heat and radioactivity for about 700

years [2]. The actinides including U also emit radioactivity and









heat during decay--for example, about 25,000 years for Pu239, 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

3X105 m3 defense HLW stored at three government sites in the United

States. These defense wastes contain 1.6X109 Curies of radioactivity

(1 Curie = 3.7X1010 disintegrations per second)[3]. There are over

2X105 m3 commercial HLW containing about 1.1X1010 Curies of

radioactivity in the form of spent fuel in the United States [31.

The total amount of HLW in Europe, Japan, the United States and

U.S.S.R. is estimated to be 10.2X105 m3 containing about 2.9X1010

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 Quan ity
Form (Ci) (m ) Source


U.S. Defense slurry 1.6X109 3X105 [3]
sludge

U.S. Commercial spent 1.1X1010 2X105 [3,41


Europe Commercial


Japan Commercial


U.S.S.R.


Total


fuel,
sludge (West
Valley, NY)

spent
fuel

spent
fuel

slurry
sludge


2.9X1 010


2X105


7X10 4


2.5X105


10.2X105


estimated


estimated


estimated


[5]












EARTH'S SURFACE kA


10-10 meters


I


INCOMING
GROUNDWATER-

DISTURBED
ZONE


-t OVERPACK
-6NASTE CANISTER
I ---
DOWNSTREAM WATER
I CONTAINING DISSOLVED
I WASTE GLASS
COMPONENTS
I


BACKFILL
BUFFER, FILLER





-- WASTE ^
oo GLASS

*'** 00s|


HOST ROCK I
(GRANITE, BASALT,
SALT, etc.


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










and nickel alloys [71, 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













Table 1-2. Candidate Waste Forms Considered for Geologic Disposal of
High Level Waste [8].


Waste Form


Comments


Borosilicate Glass


Synroc-C,D

Tailored Ceramic

High-Silica Glass

FUETAP Concrete

Coated Sol-Gel Particles

Glass Marbles in Lead
Matrix


Primary Waste Form, U.S. Reference Waste


Primary Waste Form, U.S. Reference Waste
Form

Alternative U.S. Waste Form

Semi-finalist U.S. Alternative Waste Form

Semi-finalist U.S. Alternative Waste Form

Semi-finalist U.S. Alternative Waste Form

Semi-finalist U.S. Alternative Waste Form


Semi-finalist U.S. Alternative Waste Form










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,










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 [Si04] ,

[BO1]5- and [BO3] polyhedra. Neighboring polyhedra are bonded

together by sharing strong ionic-covalent bridging oxygen bonds.





















OXYGEN
SIUCON
0 BORON
SNoU.Sr,.C
SACTINIOES
OTHER WASTE
ELEMENTS






























Fig. 1-3. Glass structure containing dissolved wastes (adapted from
[15]).









Other multivalent species such as Fe 2,+3, 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 [24,251, 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
































04) 0

E3
-1 CI









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.









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/m2, 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 400, 700 or 900C with the

ratio of glass surface area to volume of leachout (SA/V) of 0.1

cm1. 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.










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 900C for up

to 1 year ranged from 30 to 50 lm.

The long-term data obtained by Bates et al. C341 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

defines the normalized elemental mass loss NLi in g/m2:



NL. 1 (2-1)
1 f. SA
1


where mi = mass of element i in the leachate, g;

fi = mass fraction of element i in the unleached specimen,
dimensionless;

SA = specimen surface area, m2

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.









Strachan et al. [351 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















aj:Q30--~----------------------------------------
SRL 131-29.8% TDS-3A
90 C, D. 1. Water
SA/ V = 0.1 cm-'


i0o


Basalt


granite


Potential Flow Rotes in Repositories


0.1(mL/hr)
4.6'-/yr )


0.5(mL/hr) .LO(mL/yr)
23(m/yr) 46(mnyr)


LOG FLOW RATE



Plot showing the total mass loss per unit area as a
function of flow rate (adapted from [36]).


tuff


-I
IC ml. J/hr)
464(m./yr)


Fig, 2-2.


~c----.--------- --


_ I __ I _1









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 Si02 (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












TYPE I
Original glass solution interface
-\ 1 BULK---
I

Z
01
fI
J Inert gloss




DISTANCE----

TYPE III A

A1203- SiO2
/COO-FiO5
CoO P205
CoO- Si02


2 \I--BULK----

0 Dual protective
wa films on glass


;i
I c
0




I *-



iE
Ic
*0


---DISTANCE------
TYPE IV


I-BULK--
Non protective
fifm on gloss


----DISTANCE--------*


TYPE II


Selective leaching



BULK----

Protective film
on glass



--DISTANCE---

TYPE III B


TYPE V

-BULK--




I DSoluble gloss



0

DISTANCE --


The five types of glass surfaces and six surface
conditions resulting from glass-environment interactions
(adapted from [40]).


Fig. 2-3.









as 0.02 to 0.2 g/m 2.day with a time dependence of static leaching of

t-5 to t0.2 or less after 28 days at 900C.

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

[Si04] 4, [B04]5- and [B03]3- polyhedra. Since different glass

network former dissolve at different rates, selective dissolution of

matrix, structural or network species is usually observed with a

multicomponent glass containing two or more network former. 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









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 kt Even after saturation has

occurred with respect to a certain species, the glass can continue to

dissolve congruently with simultaneous precipitation of that species.










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 NSi/NLSi is used to

present the solubility limits of these elements in solutions of

different pH. The normalized solubility NSi (g/m2) is given by



NSsa (2-2)
i f. SA/V



where C = the solubility limited concentration in the leachate
at the specified conditions, g/L;

f. = the mass fraction of element i in the glass;

SA = the specimen surface area, m2;

V = the volume of the leachage, L;










NLSi = 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 230C 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 [431.

One thing that Grambow did not explain with his leach data is

why NLi/NLSi, where i is Ca, Fe, Zn, Nd or Ce, is usually larger than

1 when the solubility restrictions are removed. If NLi/(NLSi+NLB)

had been used, it may be much closer to 1, since B203 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





















01


0.01


0.001


0.0001


o.cocci


Ratio of normalized solubility to NLS (20 g/m2) for
CaCO SrCOo, Nd(OH) Fe(OH) and Zn OH)2 in MCC-1 28-day
test at 900 in solutions of different pH (adapted from
[41]).


NSI
NL,,


Fig. 2-41.

















































2 3 4 b 6 7/ 9 10
pH


11


The Si leachability of a borosilicate glass immersed in a
5-day static 230C solution buffered to various pH values
(adapted from [421).


0 015


0.010 --


0.005 F-


1 1 1 1 1 1




NETWORK NETWORK
DISSOLUTION DISSOLUTE ION




--- ION EXCHANGE -----



ITTERN WIPPB 'BS
RINE BRINE ITUF
SHALE 0TU F
GRANITE
CRITICAL pH CRiTtCAL pH


Fig. 2-5.


.










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










constituents followed by a decrease with time. Grambow et al. [47]

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

























E 80-
z -8
0 pH
S60-
Si -6
Ssat.
W 4pH
o 40 surface layer
z


i' 20

solution

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 cm /sec was assumed [46,47].









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.










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].










Burial Studies

Burial studies were started in the late 1950s and early 1960s.

Merritt and Parsons [24,251 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 [261 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

(60C in the field compared to 250C 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,










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 900C. 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 um) 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 1700C.















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, 900C and

the ambient temperature (8-100C), 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










(ABS) 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 (Si02 +

Al20 )/(R20 +20 3) wt ratio in the glasses; R20 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 900C ranged from 0.3-3 ipm/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.









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, Vaxjb, 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.





40



Table 4-1. Nominal Waste Glass Compositions (wt%) Used in the Stripa
Burial.


SRL 131 + SRL 165 + SRL 131 +
Component 29.8% TDS1 29.8% TDS1 35% TDS1 ABS 39 ABS 41 ABS 118


From glass frit
Na20 12.4 9.1 11.5 12.9 9.4 9.9
Li2O 4.0 4.9 3.7 -- 3.0 2.0
ZnO -- --- -- 3.0 2.5
MgO 1.4 0.7 1.3 -- -- --
Al203 -- -- -- 3.1 2.5 4.9
B203 10.3 7.0 9.6 19.1 15.9 14.0
Fe20 -- -- 5.7 3.6 2.9
La203 0.4 -- 0.3 -- -- --

Si02 40.6 47.7 37.6 48.5 52.0 45.5
Ti02 0.7 -- 0.7 -- -- --
ZrO2 0.4 0.7 0.3 -- -- 1.0
U02 -- -- -- 1.7 1.6 0.9

P205 -- -- -- -- -- 0.3
Cr203 -- -- -- -- 0.5
NiO -- -- -- -- -- 0.4
CaO -- -- -- -- -- 4.0


From simulated waste
Fe203 13.4 13.4 15.8 -- -- --
MnO2 3.9 3.9 4.5 0.78 0.78 0.97
Zeolite** 2.9 2.9 3.4 -- -- --
A1203 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.


SRL 131 + SRL 165 + SRL 131 +
Component 29.8% TDS1 29.8% TDS1 35% TDS1 ABS 39 ABS 41 ABS 118


Na2SO4

Cs2CO3
SrCO3
U308
Cs20O
SrO
BaO

Y203
Zr02
Mo03
Ag20
SnO
Sb20
La203
Nd203
Pr203
Ce203
CdO


100.0


0.2
0.1
0.1
1.1

























99.9


0.2
0.2
0.2

1.3

























100.0


0.89
0.26
0.46
0.15
1.29
1.65
0.01
0.02
0.004
0.72
1.22
0.38
0.76
0.03


0.89
0.26
0.46
0.15
1.29
1.65
0.01
0.02
0.004
0.72
1.22
0.38
0.76
0.03


1.11
0.33
0.58
0.19
1.62
2.06
0.01
0.02
0.005
0.90
1.53
0.48
0.95
0.03


100.0 100.0 100.1


TDS waste as received from SRL contained Fe203, Mn02, zeolite,
Al20 NiO, SiO, CaO, Na20, Coal and Na2SO4. 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.


Total










vitrification. The mixture was fused at 1150-12000C for 2-6 hours

and annealed at 500-5250C 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-im 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 um) 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.
































































































(,



(1)

r-.
aD









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-O stretching vibrations

increase with increasing Si02 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 SRL 131 + SRL 165 + SRL 131 +
(month) 29.8% TDS 29.8% TDS 35% TDS ABS 39 ABS 41 ABS 118


Minicans*
1 900C 900
3 9000C 900,
12 900C 90C
24 8-100,900C 8-100,900C


Pineapple Slices**
1 900C 900C 900C 900C 900C
2 -- -- -- -- -- 900C

3 900C 900C 900C 900C 900C
4 -- -- -- -- -- 9000C
6 90C -- -- -- -- --
7 -- -- -- -- -- 900C
12 8-100,900C 900C 900C 900C 900C 900C
24 8-100,90C 8-100,900C 8-100,900C
31 -- -- -- 90C 9000


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.













Table 4-3. Variations in Spectral Characteristics of SRL Waste
Glasses.


Peak* Peak* Integrated Area
Glass Location (cm ) Intensity (?) (Relative Value)


Minicans
SR1 131 + 29.8% TDS 989 4** 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
800-1200 cm was
** Mean and standard


containing Si-O-Si stretching vibrations at
used in the statistical analysis.
deviation.























0




O-





0- + -
So +
+o <
'Ti
OD ID
re, U) o
+ cr C13



ZC











-OM 4
0 0 0 ON
S*.. C Oa















i NV0 r n
^ S^ 0 Z










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 5000C 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












Table 4-4.


Nominal Composition
of Black Frit 165-
Mobay Glass.


Component


Si02

Fe203
Na2O

B203
Li20
Al2
A1203
Mn02
CaO
NiO
MgO
ZrO2
F
Cl
Pb
K20
Ti02
BaO
ZnO


Total


Wt %


55.61

11.34
10.44
7.23
4.82
4.22
2.11
1.10
0.90
0.70
0.90
0.14
nil
nil
0.14
0.23
0.07
0.05


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 Fe20 .

































































Fig. 4-3. A minican assembly.










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

11-2).

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


















as


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


~*- i
























































-a12j
=B~r


0-1

C)
0
0)




4,
0
CD


c0












.)
02
0)
r-4
0.


4' *














C 0)
*-4


4-,



*-4




* E

00)



-4

C OO

*-4

-rl 0)
3 C)




coi
Cr-l





inr



T-
Eb~













Table 4-5. Average Major/Minor Chemical
and Mineral Constituents in
Stripa Granite.


Oxide Wt %


SiO2 74.7
A1203 13.2
Fe20 1.6
FeO NR a
MgO 0.20
CaO 0.6
Na2O 4.0
K20 4.6
TiO2 0.05
P205 NR
MnO 0.03
BaO 0.02
H20 NR
CO2 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.










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, 900C 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 900C (Fig.

4-6) to simulate the thermal period of -300 years. Without the

heating elements, glass samples were tested at ambient mine

temperature (8-10oC), 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.


















BURIAL SITE AT STRIPA


GRANITE





MIN: 1


SHAFT


rUNNEL


~.1115

F~v ''1
1,
4 ~r~


11111 1


\ ~J


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, 900C specimens were
placed. Samples were placed in the hole to a depth of
about 2.5 m.


Fig. 4-6.


Ir
.u
1
















SCHEMATIC OF BURIAL ASSEMBLY AT STRIPA




GBANITt


i : r


,HAMiTU


I' Ni APPL I
SLIC


r ~~ trrr rr srlr
iPl Burl~!


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,
9000C assembly immediately after removal from the borehole.


Fig. 4-7.


tiKntiifliMW Lt


~ ~r


Ir~ctpyC SI1~r? R\~nililV







58














c,










O IO
0


S ca
E-











.C










bO
00
0 0


















S S O



0 C
0



0 0






0

U) V
*-*


4--





N0




-I 0



CE-
O c O




O C V


0)
- -
CO C~O









-4


r)-I O
.0 0
.O
Co a o












Table 4-7. Ground Water Composition for
the Stripa Granite, Literature
Values.


Anions
HCO3
Cl

S042-
F


mg/1
15.4-78.7
52-283
2.7-1.9
NRa


Cations
Ba2+
Ca2+
Fe3
Li
Mg2+
K+
Si02
Na+
Sr2+

pH
Total Dissolved


NR
10-59
0.02-0.24
NR
0.5
0.2-5.4
11.0-12.8
43-125
NR
8.85-9.75
200-230 (330-410 m)b

375-510 (below 700 m)b


Adapted from [60].

a NR = not reported.
Depth of sample below surface.









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 cm1 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 cleaning 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










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 900C. Samples will be

removed from the mine at time intervals of 6 months, 1 year, 2 years

and 5 years.









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 900C 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. 14-8.

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.


















































0,

S--e


C- '.
OJ




C2 /- OV
c30 /) i)


I
i

I

I


I
I
r


I

I
f
I


r










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 cm1 in the cells. In the case of

static leaching, the surface area of glass sample was about 14 cm2

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 cm2. 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











































....................
Fi4,




























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







66























O




0
0













0 D



Lu

C LLJ
>0




O











C/)







-J-







-'-I
Sr r\\







=34 l^0 \ ___ i, v-







\ | _____La










vessel was kept at 900C 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 900C 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
















Sample Matrix of the Laboratory Tests.


Glass Composition*

Temperature

SA/V

Leaching Time

Leaching Condition


Leachant


SRL 165 + 29.8% TDS, Black Frit 165-Mobay

900C

0.1 and 1.0 cm1

1, 3 and 6 months

Static and flow (0.1 and 0.3 mL/h), with and
without granite cup

deionized water
Stripa ground water
Stripa ground water saturated with glass
powders for 14 days at 900C


* Samples were run in duplicates.


Table 4-8.














RBS
SIMS
LIGHT
MICROSCOPE


5-200 A


SEM-EDS
FT-IRRS SIMS/
t t t


10.5,pm


4 I-


I.5jm


BULK GLASS


ION MILLING
SOLUTION ANALYSIS


ALTERED
LAYEF


Fig. 4-11.


Sampling depths with various techniques used in this
study (adapted from [63,64]).














Table 4-9. Characteristics of Analytical Techniques.


Sampling Spatial Detection
depth resolution Information limits(%)

-4
Secondary ion mass 5-20A 100A-=1im composition <10
spectroscopy (SIMS) (profiling structure
to =.10m

Fourier transform composition
infrared reflection structure
spectroscopy (FT-IRRS) =0.5pm 3-5mm morphology 3

Scanning electron
microscopy-energy
dispersive spectroscopy morphology
(SEM-EDS) 1.51m 1.5pm composition 5

Rutherford back
scattering (BS) 100 A- m 1mm composition >10-3
scattering (RBS) 100 A--1ym 1mm composition >10









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-Um

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

















































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 after
polishing to 600 grit surface finish (100X).









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 ym thick or greater.

Because of the 0.5-ym 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 um of the layer using near normal

specular reflectance. This nondestructive testing technique is

valuable for quick and efficient routine controls while in other










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 900C. The decoupling of the S and NS peaks in the region

of 800-1150 cm1 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, JEOL Ltd., Tokyo,
Japan.















CO

(I
0.



so
C O
0 o

rO
0 0 a)

0 4-) 0
fO --*H0 -
r- (3
S/ o a h e

0, E
// a 0a


LC 0 .0)
0 >1 0

I 0)0(0



/ l oo c-o a


.0O u o.







/ 0n ( -
a- / fl bo 3









0 L. l o 1 0
W_-


S0 0 r


SO O EN i <0
co -^ -O
^ $ i1 4->'-





H *-


S0










-r
O
(%) 30NV03-1A38Xr-A




W M
rZ4 O --

















































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


F Pl'- E































I0 >
-=w


AIISN31NI


3AllVI38










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, 900C 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































............ Si
B


S___ --Li
... *...... Ca




-CS

-- Sr


K
--- -- -i








ABS 118
GLASS/GLASS
INTERFACE, 900C
I YEAR
.2 .4 .6 I 1.4
.2 4 .6 .8 I 12 1.4


DEPTH (urm)


Fig. 4-17.


SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 900C burial in Stripa. The atomic
concentrations of all cations (except H) are summed up to
100%.


10










01




0.01





0.001










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 (4-1)
Al,after Al,before Al,abs Al,leached



where GAAl,after = gramratoms of Al after leaching;

GAl,before = gram*atoms of Al before leaching;

GAAlabs = gram-atoms of Al absorbed from solution;

GAAl leached = gram-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




GAl,after = GAl,before (42)



where GAAl,after and GAAl,before are the same as in equation (4-1).

Also, we can write












GA = at.% Z GA (4-3)
Al,after Al,after i,after



where at.%Alafter is concentration of Al (in at.%) at a certain

depth after leaching as shown in Fig. 4-17 and E GA. is a
i,after
summation of the gram-atoms of element i after leaching. Combining

equations (4-2) and (4-3) gives



GA = at.% E GA (4-4)
Al,before Al,after i. ,after



Also



GA = at.%. GA. (4-5)
i,after i,after i,after



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
GAi aaftr i,after Al,before
i,after at. ,after
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.



















































0 .2 .4 .6 .8 1 1.2 1.4
DEPTH (Jum)


Fig. 4-18.


SIMS depth profiles of ABS 118 glass/glass interface
after 1-year, 900C 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.


100





10


0


ao
a




4) E

E 0.01
00


0


0.001


.................................... ...) .. ......** Si

A .-."- Na


---E -- A'I
. _____"""""














- ABS 118
- GLASS/GLASS
_K






INTERFACE, 900 C
I YEAR
1i i i










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.% d (4-7)
1 i,bulk bulk



where i is Fe203, NiO, A1203, Zr02 or MnO2; wt%i, the wt% of the ith

oxide; d, the density of the leached layer at a specific depth, and

dbulk is the glass density. From equation (4-7),


wt%
d i,bulk (4-8)
I = (4-8)
d dbulk wt%.
bulk i


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