bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

ABSTRACTA 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.
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

ABSTRACTTwo 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

ABSTRACTSilicotitanate 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

ABSTRACTAluminosilicates 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

ABSTRACTThe 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

ABSTRACTA 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

ABSTRACTThe 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

ABSTRACTThe 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

ABSTRACTSpent 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

ABSTRACTSpent 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

ABSTRACTTo 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

ABSTRACTThe 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

ABSTRACTA 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

ABSTRACTIn 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

ABSTRACTThe 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

ABSTRACTThe 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

ABSTRACTThe 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

ABSTRACTExperiments 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.