bluuebunny/crossref_metadata_2025 huggingface dataset:
Solution-Borne Colloids from Drip Tests using actinide-Doped and
Fully-Radioactive Waste Glasses
Authors: Fortner Jeffrey A. , Wolf Stephen F. , Buck Edgar C. , Mertz Carol J. , Bates John K.
Published: September 2012

Drip tests designed to replicate the synergistic interactions between waste
glass, repository groundwater, water vapor, and sensitized 304L stainless
steel in the potential Yucca Mountain Repository have been ongoing in our
laboratory for over ten years. Results will be presented from three sets of
these drip tests: two with actinide-doped glasses, and one with a
fully-radioactive glass. Periodic sampling of these tests have revealed
trends in actinide release behavior that are consistent with their
entrainment in colloidal material when as-cast glass is reacted. Results
from vapor hydrated glass show that initially the actinides are completely
dissolved in solution, but as the reaction proceeds, the actinides become
suspended in solution. Sequential filtering and alpha spectroscopy of
colloid-bearing leachate solutions indicate that more than 80% of the
plutonium and americium are bound to particles that are captured by a 0.1 μm
filter, while less than 10% of the neptunium is stopped by a 0.1 μm filter.
Analytical transmission electron microscopy has been used to examine
particles from leachate solutions and to identify several actinide-bearing
phases which are responsible for the majority of actinide release during
glass corrosion.
Waste Glass Analytical Bias Correction Using a Reference
Standard Glass
Authors: Smith Gary L. , Eggett Dennis L. , Smith Harry D.
Published: September 2012

Analytical round robins were established to allow laboratories responsible
for the analysis of (nuclear) wastes glasses to compare their analytical
capabilities and techniques with one another in a non-competitive
atmosphere. In addition the quality of analyses using different analytical
techniques on the same materials could be compared. Analytical Round Robin 7
had two primary objectives: 1) to evaluate the effect of using the same
analytical reference glass as a reference standard for bias correcting
analytical results for all the participating laboratories in making the
analyses; and 2) to discuss the analytical methods and results from the
participating laboratories between the participants for the purpose of
evaluating quality and interlaboratory consistency. This paper will
primarily address the first objective.

The purpose of an analytical reference material is to bring the analytical
results from various laboratories into better agreement by providing them a
common reference point No reference glass is likely to be ideal for each
element in the glass being analyzed. Obviously, the closer the reference
material is to the unknown, the more confidence there will be in the results
because less judgement is called for. Use of the reference glass to bias
correct analytical results for a standard set of analyses from several
laboratories provides the data needed to determine the effectiveness of an
Analytical Reference Glass (ARG) in producing interlaboratory agreement.
Concurrently, whether or not the analytical results using ARG-1 are more
valid or systematically biased can be evaluated using standard statistical
methods. The conclusion is that bias correcting with a standard glass such
as ARG-1 for results obtained by the inductively coupled plasma - atomic
emission spectroscopy (ICP-AES) analytical technique produces more accurate
results.
Modelling Glass Dissolution in Clay with Analytic and Stochastic
Methods
Authors: Aertsens Marc
Published: September 2012

We present a stochastic method to model the dissolution of nuclear glass.
Using this method, we solve the diffusion equation in a stochastic way. We
do this by giving a large number of particles Brownian displacements.
Simultaneously, these particles can participate in other processes, like a
chemical reaction or convection.

We apply this method to solve the Pescatore model for the dissolution of
nuclear glass in clay. This model combines diffusion of silica in the pore
water of the clay with the glass dissolution rate law proposed by Grambow.
We use the model for fitting the dissolution data of four glasses in clay
slurries (with a high and with a low clay content) and in pure clay. We
present the values of the fitting parameters. The solution of the model,
obtained by the simulation method, agrees with the analytical solution. We
also extend the Pescatore model with a moving boundary, taking into account
the receding of the glass surface by corrosion.
Applicability of
V2O5-P2O5 Glass System for
Low-Temperature Vitrification
Authors: Nishi T. , Noshita K. , Naitoh T. , Namekawa T. , Takahashi K. , Matsuda M.
Published: September 2012

The V2O5-P2O5 glass system (vanadate glass) has been developed to provide
low temperature vitrification of radioactive waste without volatilization of
radionuclides. The dissolution behaviour of vanadate glass was determined
from the results of the MCC-1 static leach test at 70 ‘C using deionized
water, cement-saturated alkaline water, and cement-saturated alkaline water
with 3% NaCl. The solubility of vanadate glass was remarkably reduced by the
addition of Sb2O3to the glass. It was also lowered in
the cement-saturated water due to formation of a precipitation layer.
Although the presence of NaCl in the solution accelerated the dissolution of
vanadate glass, the addition of alkaline earth metal oxide to the glass gave
a dissolution rate three orders lower than non-additive vanadate glass. The
results of leach tests using simulated vitrified waste form containing 5moI%
of Cs2 and SrO showed there was a possibility of fixing Cs and Sr
into the glass structure.
Uranium Speciation in Glass Corrosion Layers: an XAFS Study
Authors: Biwer Bruce M. , Soderholm L. , Greegort R. B. , Lytle F. W.
Published: September 2012

Uranium L3X-ray absorption data were obtained from two borosilicate glasses, which are considered as models for radioactive wasteforms, both before and after leaching. Surface sensitivity to uranium speciation was attained by a novel application of simultaneous fluorescence and electron-yield detection. Changes in speciation are clearly discernible, from U(VI) in the bulk to (UO2)2+-uranyl in the corrosion layer. The uranium concentrations within the corrosion layer also show variations with leaching times that can be determined from the data.
Colloid Formation During the Interaction of HLW Glass with Interstitial Clay Water
Authors: Wei J. , Van Iseghem P.
Published: September 2012

During the reaction of HLW glass with interstitial clay water at different temperatures with various ratios of glass surface area to solution volume (SA/V) and durations, Eu released from the glass forms predominantly Eu-humate colloids (organic colloids) by a complexation reaction. The size distribution and stability of Eu-humate colloids have been characterized. It is likely that inorganic colloids which are mainly composed of Si, Al and Ca are generated from the corrosion of waste glass by a nucleation reaction.
Influence of the Degraded Surface Layer on the Long-Term
Behavior of Cement Pastes
Authors: Faucon P. , Jacquinot J. F. , Adenot F. , Virlet J. , Petit J. C.
Published: September 2012

Leaching of a cement paste by a demineralized solution results in the
progressive dissolution of the zone in contact with the aggressive solution.
In the long term, this dissolution will determine the kinetics of
degradation of the material. These kinetics will depend principally on the
solubility of calcium in the zone in contact with the aggressive
solution.
Science Underlying Radioactive Waste Management: Status and
Needs-Twenty Years Later
Authors: Roy Rustum
Published: September 2012

The setting of the technical subsystem within the overall
socio-political-economic-technical radwaste system will be described and the
highly interactive nature of the larger setting emphasized. It will be shown
that because of the dominance of the socio-political subsystem, the
importance of the technical subsystem is overshadowed. Moreover the key
issue in the technical subsystem, whether to put more reliance on the
immobilization (via the waste package or engineeered barriers) or the
isolation (via the geology) has stayed tilted toward the latter since 1978,
when we organized the first symposium on radwaste science (at the Materials
Research Society meeting). Now that the isolation strategy is stymied, the
opportunity arises again for the materials community to make a compelling
case for the waste package's true significance.

This review is made mainly from the perspective of one laboratory in one
country, the U.S. What is remarkable about the state of research in the
technical subsystem is how little the big picture of the
science or the technology has changed after some
billions spent on R/D. The borosilicate glass (almost unchanged) is still
the establishment's choice of reference waste form for HLW Cost, however, is
finally forcing cement encapsulated forms to be given a second look.
Mineral-modeled ceramics have received a great deal of scientific attention
but remain esoteric to managers. It will be shown that in the author's
opinion an enormous amount of detailed science has been done but most of it
is unlikely to prove to be of any relevance or use. The policy implications
for future R/D are discussed.
PCT Leach Tests of Hot Isostatically Pressed (HIPped) Zeolitic
Concretes
Authors: Siemer D. D. , Roy Delia M. , Grutzeck Michael W. , Gougar M. L. D. , Scheetz Barry E.
Published: September 2012

The logical place for the US Federal government to site a permanent
repository for its massive accumulation of cold-war reprocessing radwaste
would be at its primary cold-war nuclear device testing reservation, the
Nevada Test Site(1). Regardless of whether it eventually chooses
to implement that repository by drilling lateral “drifts” into consolidated
rock (e.g. its proposed Yucca Mountain facility) or by augering
moderately-deep boreholes into unwelded alluvium beds (e.g. the “Greater
Confinement Disposal (GCD) repository implemented at Frenchman Flats in 1984
(2), zeolitic hydroceramic materials would be more stable than
would glasses. The thermodynamic rationale for this is that in such regions,
soil solutions and soil gasses both tend to “weather” buried natural glasses
to zeolitic materials(3). It has also been demonstrated that the
same type of phases form when properly designed cementitious “grouts” are
cured under mild hydrothermal conditions(4,5) - precisely those
conditions assumed by DOE's repository modelers for “failed” (i.e. flooded)
repositories and under which radwaste-type glasses rapidly decompose to form
crystalline zeolitic phases(6,7).
Evaluation of Microbially Influenced Degradation as a Method for the Decontamination of Radio Actively Contaminated Concrete
Authors: Rogers R. D. , Hamilton M. A. , Nelson L. O. , Benson J. , Green M.
Published: September 2012

Because there are literally square kilometers of radioactively contaminated concrete surfaces within the U.S. Department of Energy (DOE) complex, the task (both scope and cost) of decontamination is staggering. Complex-wide cleanup using conventional methodology does not appear to be feasible for every facility because of prioritization, cost, and manual effort required.

We are investigating the feasibility of using microbially influenced degradation (MID) of concrete as a unique, innovative approach for the decontamination of concrete. Currently, work is being conducted to determine the practicality and cost effectiveness of using this environmentally acceptable method for decontamination of large surface concrete structures. Under laboratory conditions, the biodecontamination process has successfully been used to remove 2 mm of the surface of concrete slabs. Subsequently, initial field application data from an ongoing pilot-scale demonstration have shown that an average of 2 mm of surface can be removed from meter-square areas of contaminated concrete. The cost for the process has been estimated as $1.29/m2. Methodologies for field application of the process are being developed and will be tested. This paper provides information on the MID process, laboratory evaluation of its use for decontamination, and results from the pilot field application.
Plutonium and Neptunium Incorporation in Zirconolite
Authors: Begg B. D. , Vance E. R. , Day R.A , Hambley M. , Conradson S. D.
Published: September 2012

The incorporation of Pu and Np in zirconolite (CaZrTi2O7) has been investigated over a range of redox conditions. Zirconolite formulations designed to favour either trivalent or tetravalent Pu and Np were prepared by limiting the amount of charge compensating additives available to maintain electroneutrality. From near-edge X-ray absorption spectroscopy the Pu valence state was found to vary with the processing atmosphere, from completely tetravalent when fired in air, and located on either the Ca or Zr sites, to trivalent, when substituted on the Ca site after annealing in 3.5% H2/N2. Np was predominantly tetravalent over the range of redox conditions examined and was readily incorporated on either of zirconolite's Ca or Zr sites. The charge compensation mechanisms at work in different zirconolites are also discussed.
Synroc Derivatives for the Hanford Waste Remediation
Task
Authors: Vance E. R. , Hart K. P. , Day R. A. , Carter M. L. , Hambley M. , Blackford M. G. , Begg B. D.
Published: September 2012

Three wt% each of Cs and Tc were mixed with the standard Synroc precursor
and the ceramic was formed by hot-pressing. Attempts were made to
incorporate the Tc as either metal or Tc 4+, using different
redox conditions in processing. Volatile losses of Tc during calcination
were < 0.1% in all cases. Short-term Tc leach rates when the Tc was
present as a metal alloy were in the order of 10−4 g/m /d at 90°C
with frequently changed water, and decreased with increasing leaching time.
The valence of the Tc was monitored by X-ray absorption spectroscopy at the
drying and calcination stages of the production. The general viability of
Synroc/glass composites for immobilising the Hanford HLW sludges is further
demonstrated by using further refinements of additive schemes for the
inactive “All-blend” formulation and initial studies using the U-containing
“All-blend” waste formulation.
Fundamental Chemistry and Materials Science of Americium in
Selected Immobilization Glasses
Authors: Haire R. G. , Stump N. A.
Published: September 2012

We have pursued some of the fundamental chemistry and materials science of
americium in three glass matrices, two being high-temperature (850° and
1400°C melting points) silicate-based glasses and the third a sol-gel glass.
Optical spectroscopy was the principal investigating tool in the studies.
One aspect of this work was to determine the oxidation state exhibited by
americium in these matrices, as well as factors that control and/or may
alter this state. We have noted a correlation between the oxidation state of
the f-elements in the two high-temperature glasses with their
high-temperature oxide chemistries. One exception was americium: although
americium dioxide is the stable oxide encountered in air, when this dioxide
was incorporated into the high-temperature glasses, only trivalent americium
was found in the products. When trivalent americium was used to prepare the
sol-gel glasses at ambient temperature, and after these products were heated
in air to 800°C, again only trivalent americium was observed. Potential
explanations for the unexpected behavior of americium is offered in the
context of its basic chemistry. Experimental spectra, spectroscopie
assignments and other pertinent data obtained in the studies are
discussed.
Salt-Occluded Zeolite Waste Forms: Crystal Structures and
Transformability
Authors: Richardson J. W.
Published: September 2012

Neutron diffraction studies of salt-occluded zeolite and zeolite/glass
composite samples, simulating nuclear waste forms loaded with fission
products, have revealed complex structures, with cations assuming the dual
roles of charge compensation and occlusion (cluster formation). These
clusters roughly fill the 6–8 Å diameter pores of the zeolites. Samples are
prepared by equilibrating zeolite-A with complex molten Li, K, Cs, Sr, Ba, Y
chloride salts, with compositions representative of anticipated waste
systems. Samples prepared using zeolite 4A (which contains exclusively
sodium cations) as starting material are observed to transform to sodalite,
a denser alumi-nosilicate framework structure, while those prepared using
zeolite 5A (sodium and calcium ions) more readily retain the zeolite-A
structure. Because the sodalite framework pores are much smaller than those
of zeolite-A, clusters are smaller and more rigorously confined, with a
correspondingly lower capacity for waste containment. Details of the
sodalite structures resulting from transformation of zeolite-A depend upon
the precise composition of the original mixture. The enhanced resistance of
salt-occluded zeolites prepared from zeolite 5A to sodalite transformation
is thought to be related to differences in the complex chloride clusters
present in these zeolite mixtures. Data relating processing conditions to
resulting zeolite composition and structure can be used in the selection of
processing parameters which lead to optimal waste forms.
Comparison of Sodium Zirconium Phosphate-Structured HLW Forms
and Synroc for High-Level Nuclear Waste Immobilization
Authors: Zyryanov V. N. , Vance E. R.
Published: September 2012

The incorporation of (a) Cs and Sr as; (b) simulated actinides, and (c)
simulated Purex waste in sodium zirconium phosphate (NZP) has been studied.
The samples were prepared by sintering, by hot pressing and by hot isostatic
pressing in metal bellows containers. The short-term chemical durability of
the phosphate-based material containing Purex waste was within an order of
magnitude ofthat for Synroc-C, as measured by 7-day MCC-1 tests at 90°C. The
dissolution behaviour showed evidence of re-precipitation phenomena, even
after times as short as 28 days. Potential for improvement of NZP-based
ceramics for HLW management is discussed.
Synthetic Mineral-Like Matrices for HLLW Solidification:
Preparation by Induction Melter with a Cold Crucible (CCIM)
Authors: Smelova T. V. , Krylova N. V. , Shestoperov I. N.
Published: September 2012

A technique for solidification of high-level liquid wastes (HLLW) to obtain
mineral-like forms through the use of water-cooled melters with direct
induction heating of a melt (CCIM) has been developed at the State
Scientific Center of the Russian Federation VNIINM (SSC RF VNIINM).

Mineral-like materials of different classes such as pyroxenes,
pyrosilicates, garnets of the andradite group, titanosilicates and Synroc D
were synthesized by the CCIM method at temperatures of 1250 – 1550°C.

The preliminary X-ray diffraction studies of these materials indicate that
the structures synthesized by the CCIM method have compositions that are
identical to those of the corresponding natural analogs. The chemical
durabilities of the synthesized materials upon their exposure to distilled
water at 20°C have been determined.

In addition to the investigation of solidification of liquid HLW, the CCIM
method is examined and applied to the synthesis of mineral-like compositions
involving simulated waste components (nonsoluble residues, ashes, and pulps)
produced by the chemical and metallurgical manufacturing of fissile
materials.
Effect of Different Glass and Zeolite-a Compositions on the
Leach Resistance of Ceramic Waste Forms
Authors: Lewis M. A. , Hash M. , Glandorf D.
Published: September 2012

A ceramic waste form is being developed at Argonne National Laboratory for
waste generated during the electrometallurgical treatment of spent nuclear
fuel. The waste is generated when fission products are removed from the
electrolyte, LiCI-KCl eutectic. The ceramic waste form is a composite,
fabricated by hot isostatic pressing a mixture of glass frit and zeolite
occluded with fission products and salt. Past work has shown that the
normalized release rate (NRR) is less than 1 g/m2d for all
elements in a Material Characterization Center-Type 1 (MCC-1) leach test run
for 28 days in deionized water at 90°C (363 K). This leach resistance is
comparable to that of early Savannah River glasses. We are investigating how
leach resistance is affected by changes in the cationic form of zeolite and
in the glass composition. Composites were made with three forms of zeolite A
and six glasses. We used three-day ASTM C1220–92 (formerly MCC-1) leach
tests to screen samples for development purposes only. The leach test
results show that the glass composites of zeolites 5A and 4A retain fission
products equally well. The loss of cesium is small, varying from 0.1 to 0.5
wt%, while the loss of divalent and trivalent fission products is one or
more orders of magnitude smaller. Composites of 5A retain chloride ion
better in these short-term screens than 4A and 3A. The more leach resistant
composites were made with durable glasses that were rich in silica and poor
in alkaline earth oxides. The x-ray diffraction (XRD) results show that a
salt phase was absent in the leach resistant composites of 5A and the better
glasses but was present in the other composites with poorer leach
performance. Thus, the data show that the absence of a salt phase in a
composite's XRD pattern corresponds to improved leach resistance. The data
also suggest that the interactions between the zeolite and glass depend on
the composition of both.
Evaluation of Standard Durability Tests Towards the
Qualification Process for the Glass-Zeolite Ceramic Waste
Form
Authors: Simpson L. J. , Wronkiewicz D. J.
Published: September 2012

Glass-bonded zeolite is being developed as a potential ceramic waste form
for the disposition of radionuclides associated with the U.S. Department of
Energy's (DOE's) spent nuclear fuel conditioning activities. The utility of
several standard durability tests [e.g., Materials Characterization Center
Test #1 (MCC-1), Product Consistency Test-B (PCT-B), and Vapor Hydration
Test (VHT)] was evaluated as a first step in developing methods and criteria
that can be applied towards the process of qualifying this material for
acceptance into the DOE Civilian Radioactive Waste Management System. The
effects of pH, leachant composition, and sample
surface-area-to-leachant-volume ratios on the durability test results are
discussed, in an attempt to investigate the release mechanisms and other
physical and chemical parameters that are important for the acceptance
criteria, including the establishment of appropriate test methodologies
required for product consistency measurements.

Results from PCT-Bs conducted with 4 μm diameter salt-loaded zeolite powder
indicate that a good correlation exists between release rate and ionic size
and/or charge for the release behavior of the simulated fission products in
deionized water (DRV), EJ-13 groundwater, and brine solutions. Simulated
divalent and trivalent fission products [Sr, Ba, and rare earth (RE) ions]
were preferentially retained in the zeolite (relative to the singly ionized
cations) after tests with the salt-loaded zeolite in DIW. In general, the
preferential cation release order for salt-loaded zeolite A in DrW is Li
> Na ≥ K > Cs > Al > Si > RE > Sr > Ba. Results from
PCT-Bs with the salt-loaded zeolite A immersed in high-ionic-strength brines
at 90°C indicate a significant increase, relative to DIW tests, in the
release rates of the Sr, Ba, and RE ions despite a decrease in the release
of the Si and Al ions that make up the framework matrix of the zeolite. An
increase in the Mg and Ca concentrations in the reacted zeolites suggests
that an ion exchange process may be responsible for this increase.

Vapor hydration and MCC-1 tests were performed with ceramic waste form
monoliths of glass-bonded zeolite. The VHTs (temperatures at 120,150, and
200°C) provided useful information about the effect of glass composition on
corrosion rates and alteration phase formation, and about the overall
toughness and structural integrity of the ceramic waste form. The MCC-1 test
was investigated as an alternative to the PCT for acceptance criteria
measurements. The MCC-1 results indicate that corrosion testing with both
DIW and high-ionic-strength leachants (that specifically affect the ion
exchange behavior of the fission products) are required to fully assess the
durability of the ceramic waste form. These preliminary results establish
the utility of the MCC-1 test for providing possible acceptance criteria
measurements, including elemental release comparisons between the
environmental assessment benchmark and the ceramic waste form.
The Structure and Properties of Two New Silicotitanate
Zeolites
Authors: Balmer M. L. , Su Y. , Grey I. E. , Santoro A. , Roth R. S. , Huang Q. , Hess N. , Bunker B. C.
Published: September 2012

Two new zeolitic crystalline phases with stoichiometry,
CSTiSi2O6.5 and
CS2TiSi6O15, have been discovered.
CSTiSi2O6.5 has a crystal structure isomorphous to
the mineral pollucite, CsAlSi2O6, with Ti+4
replacing Al+3. This replacement requires a mechanism for charge
compensation. A combination of techniques including neutron diffraction,
single crystal x-ray diffraction and x-ray absorption spectroscopy have
revealed that eight extra oxygens are present per unit cell
Cs2TiSi2O6.5 as compared to pollucite.
As a result of the extra oxygen, the titanium coordination geometry is
five-fold. Pentacoordinate titanium and tetrahedral silicon form a network
structure with Cs residing in cages formed by the network. The crystal
structure of Cs2TiSi6O15 is unique, with
titanium octahedra and silicon tetrahedra forming an open framework
structure with the Cs residing in large cavities. The largest covalently
bonded ring opening to the Cs cavities in both compounds are smaller than a
Cs ion, revealing that the Cs ion has minimal mobility in the structure.
Cesium leach rates for both compounds are lower than or comparable to
borosilicate glass.
Evaluation of Cesium Sujcotitanates as an Alternative Waste
Form
Authors: Su Yali , Lou Balmer M. , Bunker Bruce C.
Published: September 2012

Silicotitanate ion exchangers are potential materials for the removal of
radioactive Cs and Sr from tank wastes. In this paper the viability of
direct thermal conversion of Cs-loaded silicotitanates to an acceptable high
level waste form has been examined. Results show that in aqueous solutions,
the Cs leach rates of crystalline silicotitanates (heat treated at 800°C)
are 0.04, 0.18, 0.4 g/m2day for Cs loadings of 1, 5, and 20 wt%,
respectively. Heating the Cs-loaded (up to 20 wt %) silicotitanates at or
above 900 °C for 1 hour further reduces the Cs leach rates to approximately
zero (beyond the lppm detection limits). Moreover, Cs volatilization was
found to be < 0.8 wt% at temperatures as high as 1000 °C. These results
suggest that thermally converted silicotitanate ion exchangers exhibit
excellent chemical durability (comparable to or better than borosilicate
glass) and thus, have great potential as an alternative waste form.
Spectroscopic Studies of Aluminosilicate Formation in Tank Waste
Simulants
Authors: Su Yali , Wang Liqiong , Bunker Bruce C. , Windisch Charles F.
Published: September 2012

Aluminosilicates are one of the major class of species controlling the
volume of radioactive high-level waste that will be produced from future
remediation at Hanford site. Here we present studies of the phases and
structures of aluminosilicates as a function of sludge composition using
X-ray powder diffraction, solid state 27Al and 29Si
NMR, and Raman spectroscopy. The results show that the content of
NaNO3 in solution has significant effects on the nature of the
insoluble aluminosilicate phases produced. It was found that regardless of
the initial Si:Al ratio, nitrate cancrinite was the main phase formed in the
solution with pH of 13.5 and 5 MNaNO3. However, at lower
NaNO3 concentration with initial Si:Al ratios of 1.1, 2.2, and
11.0 in the solutions, a range of aluminosilicate zeolites was produced with
Si: Al ratios of 1.1, 1.3, and 1.5, respectively. Lowering the solution pH
appears to promote the formation of amorphous aluminosilicates. The results
presented here are important for the prediction of the solubility and
dissolution rate of Al in tank wastes.
The Effects of Surface Modification on the Speciation of Metal
Ions Intercalated into Aluminosilicates
Authors: Wasserman Stephen R. , Giaquinta Daniel M. , Yuchs Steven E. , Soderholm L.
Published: September 2012

The effect of the addition of an organic monolayer to the surface of a clay
mineral on the speciation of metal ions intercalated into the clay
interlayer is probed by X-ray absorption spectroscopy. The presence of the
monolayer changes the surface of the clay from hydrophilic to hydrophobic.
It inhibits the interlayer ions from exchanging freely into environmental
water and reduces the leach rate of cations out of the clay by approximately
a factor of 20. Significant changes are observed when these coated samples
are treated under hydrothermal and thermal conditions. Reductions of
uranium(VI), in the form of uranyl, and copper(II) ions occur. In addition,
the uranium aggregates, forming small particles that appear similar to
UO2. Comparable conglomeration occurs with lead cations and
with the reduced copper species.
A Kinetic Model for the Stability of the Spent Fuel Matrix Under
Oxic Conditions: Model Development Against Experimental
Evidence
Authors: Bruno Jordi , Cera E. , Duro L. , Eriksen T. E. , Sellin P. , Spahiu K. , Werme L. O.
Published: September 2012

A kinetic model recently developed [1] for the radiolytically induced
oxidative dissolution of the spent fuel matrix is presented. This is based
on experimental studies on the generation and evolution of radiolytic
products in a closed system containing fragments of PWR-fuel [2]. The
outcome of this model is currently being integrated in the present PA
exercise being prepared by SKB. The calibration of the model against various
experimental information and its predictive capabilities for the long term
performance of the spent fuel matrix are presented.
Effect of Temperature and Bicarbonate Concentration on the
Kinetics of UO2(s) Dissolution Under Oxidizing
Conditions
Authors: de Pablo J. , Casas I. , Giménez J. , Molera M. , Torrero M. E.
Published: September 2012

The dissolution rate of unirradiated UO2 (s) has been studied as
a function of hydrogen carbonate concentration at three different
temperatures (298.15 K, 313.15 K and 333.15 K) under oxidizing conditions in
a continuous flow-through reactor with a thin layer of solid particles
(particle size from 100 to 300 μm). From the results of these experiments,
two different rate laws have been determined. At high temperature (313.15 K
and 333.15 K), we obtained a dissolution rate proportional to hydrogen
carbonate concentration while at 298.15 K, the rate almost depends on the
square root of the hydrogen carbonate concentration. This indicates a
different reaction mechanism depending on temperature which can be related
to the oxidation step of the overall process. The apparent activation energy
obtained was 41 kJ mo1−1.
Comparison and Modeling of Aqueous Dissolution Rates of Various
Uranium Oxides
Authors: Steward S. A. , Mones E. T.
Published: September 2012

The purpose of this work has been to measure and model the intrinsic
dissolution rates of uranium oxides under a variety of well-controlled
conditions that are relevant to a geologic repository. When exposed to air
at elevated temperature, spent fuel may form the stable phase
U3O8. Dehydrated schoepite,
UO3H2O, has been shown to exist in drip tests on
spent fuel.

Equivalent sets of U3O8 and
UO3H2 dissolution experiments allowed a systematic
examination of the effects of temperature (25–75°C), pH (8–10) and carbonate
(2–200×10−4 molar) concentrations at atmospheric oxygen
conditions.

Results indicate that UO3H2O has a much higher
dissolution rate (at least ten-fold) than U3O8 under
the same conditions. The intrinsic dissolution rate of unirradiated
U3O8 is about twice that of UO2.
Dissolution of both U3O8 and
UO3.H2O shows a very high sensitivity to carbonate
concentration. Present results show a 25 to 50-fold increase in
room-temperature UO3H2O dissolution rates between the
highest and lowest carbonate concentrations.

As with the UO2 dissolution data the classical observed chemical
kinetic rate law was used to model the U3O8
dissolution rate data. The pH did not have much effect on the models, in
agreement with the earlier analysis of the UO2 and spent fuel
dissolution data,. However, carbonate concentration, not temperature, had
the strongest effect on the U3O8 dissolution rate. The
U3O8 dissolution activation energy was about 6000
cal/mol, compared with 7300 and 8000 cal/mol for spent fuel and
UO2 respectively.
Modelling Oxidative Dissolution of Spent Fuel
Authors: Neretnieks Ivars
Published: September 2012

Spent nuclear fuel will, by the radiation, split nearby water into oxidizing
and reducing compounds. The reducing compounds are mostly hydrogen that will
diffuse away. The remaining oxidizing compounds can oxidize the uranium
oxide of the fuel and make it more soluble. The oxidised uranium will
dissolve and diffuse away. The nuclides previously incorporated in the spent
fuel matrix can then be released and also migrate away from the fuel.

A model is proposed where the produced oxidizing species compete for
reaction with the fuel and for escaping out of the system. The chemical
reaction rate of oxygen and fuel is taken from literature values based on
experiments. The escape rate of oxidants to a receding redox front in the
backfill is modelled assuming a redox reaction of oxidizing component and
reducing component in the surrounding. The rate of movement of the redox
front is determined from the rate of production of oxidants. This is
estimated using a previously devised model that has been calibrated to in
situ observed radiolysis.

Three cases are modelled. In the first case it is assumed that the reducing
compound is insoluble and that the reaction between oxygen and reducing
mineral is very fast. In the second case it is assumed that the reducing
component has a known solubility and that it can migrate to meet the oxygen
and quickly react. In a third case a finite reaction rate is modelled
between the oxygen and the reducing species.

The sample calculations show that if the reducing mineral has to be supplied
from the backfill a large fraction of the spent fuel could be oxidised. If
the corrosion products of a degraded steel canister can supply the reducing
species and the redox reaction is fast, very small amounts of the fuel could
be oxidised. Literature data indicate that the redox reaction rate may not
be so fast that it can be considered instantaneous and then a considerable
fraction of the fuel could be oxidised. The model gives a means of exploring
which mechanisms and data may be of most importance for radiolytic fuel
dissolution, but the realism of the data and the model must be tested
further. There is a lack of understanding and data on reaction rates,
heterogeneous as well as homogeneous. This is crucial to the results.
Transuranium Element Incorporation into the
β-U3O8 Uranyl Sheet
Authors: Miller M. L. , Burns P. C. , Finch R. J. , Ewing R. C.
Published: September 2012

Spent nuclear fuel (SNF) is unstable under oxidizing conditions. Although
recent studies have determined the paragenetic sequence for uranium phases
that result from the corrosion of SNF, there are only limited data on the
potential of alteration phases for the incorporation of transuranium
elements. The crystal chemical characteristics of transuranic elements (TUE)
are to a certain extent similar to uranium; thus TUE incorporation into the
sheets of uranyl oxide hydrate structures can be assessed by examination of
the structural details of the β-U3O8 sheet type.

The sheets of uranyl polyhedra observed in the crystal structure of
β-U3O8 also occur in the mineral billietite
(Ba[(UO2)3O2(OH)3]2(H2O)4),
where they alternate with α-U3O8 type sheets.
Preliminary crystal structure determinations for the minerals ianthinite,
([U24+(HO2)4O6(HO)4(H2O)4](H2O)5),
and “wyartite II” (mineral name not approved by IMA committee on mineral
names),
{CaCo3}[U4+(UO2)2O3(OH)2](H2O)4,
indicate that these phases also contain β-U3O8 type
sheets. The β-U3O8sheet anion topology contains
triangular, rhombic, and pentagonal sites in the proportions 2: 1:2. In all
structures containing β-U3O8 type sheets, the
triangular sites are vacant. The pentagonal sites are filled with
U6+O2 forming pentagonal bipyramids. The rhombic
dipyramids filling the rhombic sites contain U6+O2 in
billietite, U4+O2 in
β-U3O8U4+(H2O)2 in
ianthinite, and U4+O3 in “wyartite-II” (in which one
apical anion is replaced by two O atoms forming a shared edge with a
carbonate triangle of the interlayer). Interlayer species include:
H2O (billietite, “wyartite II”, and ianthinite),
Ba2+ (billietite) Ca2+ (”wyartite II”), and
CO3−2 (”wyartite II”); there is no interlayer in
β-U3O8. The similarity of known TUE coordination
polyhedra with those of U suggests that the β-U3O8
sheet will accommodate TUE substitution coupled with variations in apical
anion configuration and interlayer population providing the required charge
balance.
Method of Treatment of Radioactive Silts and
Soils
Authors: Varlakov A. P. , Sobolev I. A. , Barinov A. S. , Dmitriev S. A. , Karlin S. V. , Flit V. U.
Published: September 2012

To solve the problem of silts and soils that are contaminated with
radioactive and toxic substances, the following method has been developed at
SIA RADON. The material is mixed with limestone and other components,
including up to 70 % (mass) of dried residue of liquid radioactive waste.
The mixture is heated at 800 to 1000 °C, shredded, and used to form cement.
This cementation process may be used to treat radioactive or other chemical
waste.

The work demonstrated that in the case of silts, for example, the product
volume is reduced by a factor of 1.5 to 3 compared to the initial silts
volume. A fast hardening, durable product is obtained, the quality of which
is not inferior to that obtained by using the traditional binders. For some
parameters, e.g., hardening rate and macroelement leaching, the current
method is significantly better than the traditional cementation process. It
has also been demonstrated that, over a wide range of parameters, when dry
residue of liquid radioactive waste is used in the initial mixture, the part
of NOxCan be reduced to nitrogen, thereby reducing NOx
emissions.
Radiation Stability of Crown Ether Used for Decontamination of
Radionuclides from HLW
Authors: Abashkin Vyacheslav M. , Egorov Gennadiy F. , Filippov Evgeniy , Nardova Anna K. , Mamakin Igor V.
Published: September 2012

The gamma-ray initiated oxidation of dicyclohexano-18-crown-6 (DCH-6) was
studied in two-phase system under agitation: nitric acid, aqueous solution
and organic solvent.

The distribution coefficients of radionuclides, hydrodynamic characteristics
(phase separation and interface surface tension), and the radiolysis
products after irradiation at a dose rate (60Co) of 1.7 Gy/s were
determined. The maximum total dose was 3.O×105 Gy
(3.0×107R).
Plasma Chemical Processing of Depleted Uranium
Hexafluoride
Authors: Ivanov A. V. , Abashkin V. M. , Filippov E. A. , Tumanov Y. N.
Published: September 2012

A reprocessing of uranium hexafluoride waste, which was accumulated in the
nuclear industry for many years, is a very vital question now. There are a
lot of gas tanks filled with UF6 that have been exposed for a
long time and are posing a serious ecological danger. A plasma chemical
technique is proposed here to transform the uranium hexafluoride depleted
with isotope 235U (less than 0.1%) into the safe storage form -
powdered uranium oxides. The second product is HF (70–80% aqueous solution)
which may be easily reprocessed by means of the rectification column up to
pure anhydrous hydrogen fluoride.

The designed technological scheme and the basic technical characteristics of
the plasma chemical converter are presented. The proposed technique is
efficient, easy controlled and free of any harmful by-products.
Molecular Dynamics Simulation of the Trajectory of a Recoil
Nucleus in a Simplified Nuclear Glass
Authors: Delaye J. M. , Ghaleb D.
Published: September 2012

In a simplified (SiO2, B2O3,
Na2O, Al2O3, ZrO2) glass,
corresponding to the basic matrix for the French nuclear waste containment
glass, the authors developed a molecular dynamics simulation of the atom
displacement cascades resulting from a disintegration of the actinides.
After simulating the secondary cascades resulting from the first atoms
displaced by a collision with a recoil nucleus, an actinide (U) was added to
the model and the cascade produced by accelerating this type of atom was
explicitly investigated, notably to observe the morphological evolution of
the cascades and the resulting changes in the glass structure.
Effect of Radiation on Topopah Spring Tuff Mechanical
Properties
Authors: Berge P. A. , Blair S. C.
Published: September 2012

The effect of radiation on the mechanical properties of Topopah Spring tuff
was investigated by performing uniaxial compressive tests on irradiated and
control samples of the tuff from the potential repository horizon at Yucca
Mountain. Test results are presented, including stress-strain curves and
peak strength and Young's modulus values. The results from this preliminary
study show that for uncracked samples of Topopah Spring tuff, exposure to
gamma radiation had no discernible effect on the unconfined compressive
(peak) strength or the Young's modulus. However, results for samples that
contained partially healed subvertical cracks indicate that exposure to
radiation may reduce the strength and Young's modulus significantly. This is
attributed to weakening of the cementing materials in the cracks and
fractures of the samples that were irradiated. These results are
preliminary, and additional studies are warranted to evaluate whether
radiation weakens cementing materials in welded tuff.
Transmission Electron Microscopy Study of α-Decay Damage in
Aeschynite and Britholite
Authors: Gong W. L. , Wang L. M. , Ewing R. C. , Chen L. F. , Lutze W.
Published: September 2012

The aeschynite structure-type
(Ce,Nd,La,Th,U,Ca)(Nb,Ti)2O6, and the rare-earth
silicate apatite structure-type with the formula
(Ce,La,Nd,Ca,Th)10(SiO4,PO4)6(O,F,OH)2
are important rare-earth and actinide host phases for high-level nuclear
waste. Natural phases of these structure-types have calculated alpha-decay
doses up to ∼1017 α-events/mg which have accumulated over
hundreds of millions of years. Transmission electron microscopy has been
used to study the microstructure of α-decay damage in aeschynite and
britholite. Electron diffraction analysis of natural aeschynite revealed
that minerals originally crystalline gradually lost their crystallinity with
increasing alpha-decay doses. Helium bubbles were found in the aeschynite
which have accumulated up to ∼2×1016 α-events/mg. These bubbles
may nucleate within collision cascades during a-decay damage. Electron
irradiation has an enhanced rare-gas migration and the formation of larger
bubbles. High-resolution electron microscopy (HRTEM) revealed that
amorphization during accumulation of a-decay damage was from alpha-recoil
nuclei collision cascades, in both the aeschynite and britholite.
The Corrosion Behavior of Iron and Aluminum Under Waste Disposal
Conditions
Authors: Fujisawa R. , Cho T. , Sugahara K. , Takizawa Y. , Horikawa Y. , Shiomi T. , Hironaga M.
Published: September 2012

The generation of hydrogen gas from metallic waste in corrosive disposal
environment is an important issue for the safety analysis of low-level
radioactive waste disposal facilities in Japan. In particular iron and
aluminum are the possibly important elements regarding the gas generation.
However, the corrosion behavior of these metals has not been sufficiently
investigated under the highly alkaline non-oxidizing disposal conditions
yet.

We studied the corrosion behavior of iron and aluminum under simulated
disposal environments. The quantity of hydrogen gas generated from iron was
measured in a closed cell under highly alkaline non-oxidizing conditions.
The observed corrosion rate of iron in the initial period of immersion was 4
nm/year at 15 °C, 20 nm/year at 30 °C, and 200 nm/year at 45 °C. The
activation energy was found to be 100 kJ/mol from Arrhenius plotting of the
above corrosion rates.

The corrosion behavior of aluminum was studied under an environment
simulating conditions in which aluminum was solidified with mortar. In the
initial period aluminum corroded rapidly with a corrosion rate of 20
mm/year. However, the corrosion rate decreased with time, and after 1,000
hours the rate reached 0.001 to 0.01 mm/year.

Thus we obtained data on hydrogen gas generation from iron and aluminum
under the disposal environment relevant to the safety analysis of low-level
radioactive disposal facilities in Japan.
Assessing the Effect of Dissolved Organic Ligands on Mineral
Dissolution Rates: an Example from Calcite Dissolution
Authors: DeMaio Teri , Grandstaff D. E.
Published: September 2012

Experiments suggest that dissolved organic ligands may primarily modify
mineral dissolution rates by three mechanisms: (1) metal-ligand (M-L)
complex formation in solution, which increases the degree of
undersaturation, (2) formation of surface M-L complexes that attack the
surface, and (3) formation of surface complexes which passivate or “protect”
the surface. Mechanisms (1) and (2) increase the dissolution rate and the
third decreases it compared with organic-free solutions. The types and
importance of these mechanisms may be assessed from plots of dissolution
rate versus degree of undersaturation.

To illustrate this technique, calcite, a common repository cementing and
vein-filling mineral, was dissolved at pH 7.8 and 22°C in
Na-Ca-HCO3-Cl solutions with low concentrations of three
organic ligands. Low citrate concentrations (50 μm) increased the
dissolution rate consistent with mechanism (1). Oxalate decreased the rate,
consistent with mechanism (3). Low phthalate concentration (< 50 μM)
decreased calcite dissolution rates; however, higher concentrations
increased the dissolution rates, which became faster than in inorganic
solutions. Thus, phthalate exhibits both mechanisms (2) and (3) at different
concentrations. In such cases linear extrapolations of dissolution rates
from high organic ligand concentrations may not be valid.
Liquidus Temperature of Spinel Precipitating High-Level Waste
Glasses
Authors: Mika M. , Schweiger M. J. , Vienna J. D. , Hrma P.
Published: September 2012

The liquidus temperature (TL) often limits the loading of
high-level waste in glass through the constraint that TL must be
at least 100°C below the temperature at which the glass viscosity is 5 Pa-s.
In this study, values of TL for spinel primary crystalline phase
were measured as a function of glass composition. The test glasses were
based on high-iron Hanford Site tank wastes. All studied glasses
precipitated spinel (Ni,Fe,Mn)(Cr,Fe)2O4 as the
primary crystalline phase. TL was increased by additions of
Cr2O3, NiO, Al2O3,
Fe2O3, MgO, and MnO; while Li2O,
Na2O, B2O3, and SiO2 had a
negative effect. Empirical mixture models were fitted to data.