bluuebunny/crossref_metadata_2025 huggingface dataset:
Syntheses of Titanate-Based Hosts for the Immobilization of
Pu(III) and Am(III)
Authors: Shoup Shara S. , Bamberger Carlos E.
Published: September 2012

Syntheses of novel titanate-based waste forms, potentially suitable for
immobilizing Pu(III) and Am(III), have been performed. The dititanate
compounds An2Ti2O7 (An=Pu or Am) were found
to have a monoclinic structure. A series of pyrochlore-type solid solutions
between An2Ti2O7 and the cubic pyrochlore
Ln2Ti2O7 (Ln=Gd, Er, or Lu) has been
established, and limits of solid solubility for
An2Ti2O7 were determined. These limits
were found to increase as the ionic radius of the lanthanide in the host
decreased. Strontium-containing titanate compounds
SrAn2Ti4O12 were also prepared and were
characterized as exhibiting a perovskite-type structure. Such compounds
could serve as waste hosts for the immobilization of not only the actinide
but also 90 Sr. A 2 month leaching study of
Er1.78Pu0.22Ti2O7 and
SrPu2Ti4O12 by WIPP “A” brine was
performed. Very low amounts (< 1 ppm) of Pu(III) were leached. Although
the ionic radius of Np(III) is similar to that of Pu(III) and Am(III),
analogous Np compounds could not be prepared in any of the titanate systems
investigated.
Preparation of Monophasic [NZP] Radiophases: Potential Host
Matrices for the Immobilization of Reprocessed Commercial High-Level
Wastes
Authors: Hawkins H. T. , Scheetzj B. E. , Guthrie G. D.
Published: September 2012

The compositional flexibility of the sodium zirconium phosphate
(NaZr2(PO4)3) structure has been
exploited in the design of monophasic radiophases capable of immobilizing
the most common cations associated with reprocessed high-level commercial
waste streams. Highly crystalline, monophasic members of the
NaZr2(PO4)3 structural family ([NZP])
have been prepared with conventional processing methods and equipment. These
radiophases were tailored to accommodate 10–20 wt % modified PW-4b simulated
calcine as single phases isostructural with
NaZr2(PO4)3. To meet the challenge of
designing monophasic materials capable of accommodating the chemical
complexity of PW-4b, an ionic substitution scheme based on crystal chemical
principles was developed. The radiophases were prepared with inexpensive,
inorganic precursors and a solution sol-gel method; these materials were
heat treated and/or sintered under a variety of conditions to determine the
optimum conditions for single phase [NZP] formation. X-ray powder
diffraction provided valuable information that was used to assess the
suitability of the ionic substitution model developed in this investigation.
The results of this investigation suggest that monophasic [NZP] radiophases
capable of accommodating 10–20 wt % modified PW-4b simulated calcine may be
continuously processed with conventional ceramic processing methods and
equipment. Moreover, the relatively low temperatures involved and the
reproducibility of the process make [NZP] radiophases economically
attractive hosts for radioactive and heavy metal industrial wastes.
Phase Equilibria and Elements Partitioning in Zirconolite-Rich
Region of Ca-Zr-Ti-Al-Gd-Si-O System
Authors: Knyazev O. A. , Stefanovsky S. V. , Ioudintsev S. V. , Nikonov B. S. , Omelianenko B. I. , Mokhov A. V. , Yakushev A. I.
Published: September 2012

Zirconolite-rich ceramics were produced by the cold crucible melting
technique in an air atmosphere, at 1550±50 °C and 1 atm. Four samples with
overall composition (in wt.%): 4.9–14.3 CaO; 19.0–41.3 ZrO2;
24.1–42.6 TiO2; 1.3–11.3 Al2O3; 6.8–30.0
Gd2O3, and 1.1–8.5 SiO2 have been
studied. Total phases in the ceramics consist of major zirconolite and minor
rutile, perovskite, zirconia, aluminium titanate, and glass. The
Gd2O3 content in zirconolite reaches up to 31.4
wt.% corresponding to the formula:
(Cao4,Gd0.7)Zri.0(Tii4,Alo 5)070. The
data on the phase composition agree well with coupled Gd incorporation into
the mineral structure. Ca(H) + Ti(IV) = Gd(III) + Al(III), and 2Gd(III) -
Ca(II) + Zr(IV). The highest Gd contents observed in the other phases are
25.4 % for zirconia, 12.6 % in glass, 8.8 % in perovskite, and 1.4 % for
rutile. The rest of the elements' distribution in the samples are
analyzed.
Chemical Durability Study of Synroc-C Ceramics Produced by
Through-Melting Method
Authors: Kudrin A. V. , Nikonov B. S. , Stefanovsky S. V.
Published: September 2012

An interaction between Synroc-C samples prepared by inductive melting in a
cold crucible and deionized water was investigated at room temperature, 100
and 200 °C. The leachability of powdered specimens with grainsize 0.10–0.15
mm was determined by repeated static tests with regular replacement of
leachant. The experimental data showed that the most leachable elements are
the alkali, alkaline earths and molybdenum. The matrix elements such as
titanium, zirconium, and rare earths as well are very leach resistant.
Comparison of the calculated leach rates of components with reference data
on leaching of alternative materials showed that cesium leachability from
the studied specimens are close to borosilicate glass, but leaching behavior
of other components is comparable to leachability from Synroc-C prepared by
hot-pressing.
Spectroscopic Investigations of Neptunium's and Plutonium's
Oxidation States in Sol-Gel Glasses as a Function of Initial Valence and
Thermal History
Authors: Stump N. A. , Haire R. G. , Dai S.
Published: September 2012

Several oxidation states of neptunium and plutonium, Pu(III), Pu(IV),
Pu(VI), Np(IV), Np(V), and Np(VI), were studied in glasses prepared by a
sol-gel technology. The oxidation state of these actinides in the sol-gel
product was examined by absorption spectroscopy after solidification, aging,
and thermal treatment. The oxidation state of the actinides in the starting
solutions was essentially maintained through the solidification process of
the silica matrix. However, during densification and removal of residual
solvents at elevated temperatures, both actinides converted eventually to
their tetra valent states while in the different sol-gel products. This
finding is in accord with reports that tetravalent states of plutonium and
neptunium are acquired in glass products prepared by dissolution of the
actinide in molten glasses. Comparisons between room temperature spectra
obtained from neptunium and plutonium in heated sol-gel products and from
molten glass products showed subtle differences that can be related to the
metal ion's environments.
The Development of a Ceramic Waste Form for Immobilisation of
Highly Active Wastes from Radical Purex Reprocessing
Operations
Authors: Maddrell Ewan
Published: September 2012

The development of novel, Radical Purex, reprocessing technologies, leading
to fission product waste streams with high levels of inert constituents, may
mean that vitrification is no longer the optimum solution for the
immobilisation of highly active wastes at the back end of the nuclear fuel
cycle. A ceramic phase assemblage is described which uses the inert
constituents of the waste stream as a functional component of the waste
form, permitting high waste loadings to be achieved and thus enabling waste
minimisation considerations to be satisfied. The initial development of this
phase assemblage is presented.
Determination of Cl Impurities and 36Cl Instant
Release from Used Candu Fuels
Authors: Tait J. C. , Cornet R. J. J. , Chant L. A. , Jirovec J. , McConnell J. , Wilkin D. L.
Published: September 2012

Chlorine-36 has been identified as a potential source of radiological risk
in the disposal of nuclear fuel waste. The radioisotope 36Cl
(t1/2 = 3 × 1O5 a) is produced by neutron
activation of Cl impurities in UO2 fuel. The total average Cl
impurity level in four unirradiated CANDU UO2 fuel samples was
2.3 ± 1.1 ppm. ORIGEN-S calculations using a 5 ppm Cl impurity in a CANDU
fuel resulted in a 36Cl activity comparable to the activity of
129I and 14C produced in the fuel thus requiring
36Cl to be considered in disposal risk assessments. The
“instant release” of 36Cl from the gap and grain boundary regions
of the fuel to solution was measured by leaching both clad fuel and fuel
samples crushed to grain-sized particles. The 36Cl concentration
was measured by Accelerator Mass Spectrometry. The 36Cl releases
from fuel samples taken from 8 different fuel bundles ranged from 0.5% to
20.4% of the total 3 Cl inventory over a leaching period of 32
days. The 36Cl released was found to correlate with the stable Xe
gas release, the fuel burnup and the linear power rating (LPR). For a
typical CANDU fuel with an LPR of -42 kW/m, the “instant release” of
36Cl would be about 5% of the total inventory.
Leaching of Used Candu Fuel: Results from a 19-Year Leach Test
Under Oxidizing Conditions
Authors: Stroes-Gascoyne S. , Johnson L. H. , Tait J. C. , McConnell J. L. , Porth R. J.
Published: September 2012

A fuel leaching experiment has been in progress since 1977 to study the
dissolution behaviour of used CANDU fuel in aerated aqueous solution. The
experiment involves exposure of 50-mm clad segments of an outer element of a
Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53
and 58 kW/m, respectively), to deionized distilled water (DDH2O,
∼2 mg/L carbonate) and tapwater (∼50 mg/L carbonate). In 1992, it was
observed that the fuel in at least one of the leaching solutions showed some
signs of deterioration and, therefore, in 1993, parts of the fuel samples
were sacrificed for a detailed analysis of the physical state of the fuel,
using SEM and optical microscopy. Leaching results to date show that even
after >6900 days only 5 to 7.7% of the total calculated inventory of
137Cs has leached out preferentially and that leach rates
suggest a development towards congruent dissolution. Total amounts of
137Cs and 90Sr leached are slightly larger in
tapwater than in DDH2O. SEM examinations of leached fuel surface
fragments indicate that the fuel surface exposed to DDH2O is
covered in a needle-like precipitate. The fuel surface exposed to tapwater
shows evidence of leaching but no precipitate, likely because uranium is
kept in solution by carbonate. Detailed optical and SEM microscopy
examinations on fuel cross sections suggest that grain-boundary dissolution
in DDH2O is not prevalent, and in tapwater appears to be limited
to the outer %0.5 mm (pellet/cladding) region of the fuel. Grain boundary
attack seems to be limited to microcracks at or near the surface of the
fuel. It thus appears that grain-boundary attack occurs only near the fuel
pellet surface and is prevalent only in the presence of carbonate in
solution.
Grain Boundary Corrosion and Alteration Phase Formation During
the Oxidative Dissolution of UO2 Pellets
Authors: Wronkiewicz David J. , Buck Edgar C. , Bates John K.
Published: September 2012

The alteration behavior of UO2 pellets following their reaction
under unsaturated drip-test conditions, at 90°C, for time periods of up to
10 years has been examined by solid phase and leachate analyses. Sample
reactions were characterized by preferential dissolution of grain boundaries
between the original press-sintered UO2 granules comprising the
samples, development of a polygonal network of open channels along the
intergrain boundaries, and spallation of surface granules that had undergone
severe grain boundary corrosion. The development of a dense mat of
alteration phases after two years of reaction trapped loose granules,
resulting in reduced rates of particulate uranium release. The paragenetic
sequence of alteration phases that formed on the present samples was similar
to that observed in surficial weathering zones of natural uraninite
(UO2) deposits, with alkali and alkaline earth uranyl
silicates representing the long-term solubility-limiting phases for uranium
in both systems.
Coprecipitation Studies from Simfuel Solutions in 5 m NaCl
Media
Authors: Diaz-Arocas P. , Garcia-Serrano J.
Published: September 2012

Extensive Research is performed in many countries in order to evaluate the
spent fuel behaviour under repository conditions. Several aspects as the
control of the oxidative spent fuel dissolution by secondary phases
formation are not yet clear.

Coprecipitation experiments from SIMFUEL solutions are performed to study if
minor elements will influence the formation of secondary phases. Therefore,
coprecipitation studies from SIMFUEL solutions aims at identification of
stable phases of significant simulated fission products. These experiments
provide upper limits for solution concentration and distribution ratios of
simulate fission products at several pH values. SIMFUEL pellets, which
simulate an irradiated fuel with burnup of 50 GWd/tU were provided by AECL
Research Laboratories, Canada. Experiments were carried out by addition of
an aliquot of the initial SIMFUEL solution to 5 m NaCI free of carbonates
solution. The selected pH was maintained constant during the experiments.
The pH range considered was from 5.5 to 9.3. Analyses of the solutions were
performed for uranium by Laser fluorescence and for the minor elements by
ICP-MS. Solid phases formed at pH 5.5 were dissolved and analysed by ICP-MS.
Results of the evolution in solution vs. pH of simulated fission products
concentrations are shown in this paper.
Use of Induction Melter with a Cold Crucible (CCIM) For HLLW and
Plutonium Immobilization
Authors: Kushnikov V. V. , Matyunin Yu. I. , Smelova T. V. , Demin A. V.
Published: September 2012

CCIM technology allows syntheses of various materials over a wide range of
compositions. This provides a means of applying the CCIM to the
solidification of wastes obtained by different technologies of the spent
nuclear fuel reprocessing, including toxic wastes containing heavy metals
and those arising from fractionation. Solid compounds produced by this
method meet both the requirements for their subsequent disposal in deep
geological formations and spent fuel standards.

On the basis of investigations into the REE and Uranium vitrification, the
possibilities of CCIM immobilizing weapons Plutonium in preparation of glass
compositions are considered.
Dissolution Studies of Soddyite as a Long-Term Analogue of the
Oxidative Alteration of the Spent Nuclear Fuel Matrix
Authors: Pérez I. , Casas I. , Torrero M. E. , Cera E. , Duro L. , Bruno J.
Published: September 2012

The thermodynamic and kinetic dissolution properties of a synthetically
obtained soddyite have been determined at different bicarbonate
concentrations. This uranium-silicate is expected to be a secondary solid
phase of the oxidative alteration pathway of uranium dioxide in waters with
low phosphate content and, consequently, it is likely to constitute one of
the long-term uranium solubility limiting solid phases.

The experimental data obtained at the end of the experiments correspond
fairly well to the theoretical model calculated with a log K0S0 of 3.9±0.7.

On the other hand, the general trend of the total uranium in solution
measured in the experiments as a function of time has been fitted by using a
kinetic equation obtained from the principle of detailed balancing of the
dissolution reaction. In addition, the EQ3/6 code has also been used to
model the uranium concentrations as a function of time. In both modeling
exercises comparable results were obtained. The dissolution rate, normalized
to the total surface area used in the experiments as measured with the BET
method, gave an average value of 6.8 (±4.4) 10−14 mol
cm−2 s−1.
New Sorbents and Technologies for Liquid Radioactive Waste
Management
Authors: Marinin D. V. , Avramenko V. A. , Glushchenko V.Yu. , Sergienko V.l. , Vasilevskiy V. A. , Zheleznov V. V.
Published: September 2012

The possibilities for using various technologies for liquid radioactive
waste (LRW) management were considered. Technologies based on the
application of sorbents highly selective to radionuclides were shown to have
good prospects. Possible ways of increasing selective sorbent efficiency in
cleaning radionuclide-polluted waters of various types were examined. The
results of studying new high-selective sorbents produced in the Institute of
Chemistry FEDRAS are given. The use of these sorbents enables one to treat
LRW of various types in a single stage. The results of testing a sorption
filter installation for management of LRW produced during operation, repair
and disposal of nuclear reactors are presented. Use of this installation
resolves a major portion of LRW disposal problems.
Depleted Uranium Oxides and Silicates as Spent Nuclear Fuel
Waste Package Fill Materials
Authors: Forsberg C. W.
Published: September 2012

A new repository waste package (WP) concept for spent nuclear fuel (SNF) is
being investigated. The WP uses depleted uranium (DU) to improve performance
and reduce the uncertainties of geological disposal of SNF. The WP would be
loaded with SNF. Void spaces would then be filled with DU (∼0.2 wt %
235U) dioxide (UO2) or DU silicate-glass beads.

Fission products and actinides can not escape the SNF UO2
crystals until the UO2 dissolves or is transformed into other
chemical species. After WP failure, the DU fill material slows dissolution
by three mechanisms: (1) saturation of WP groundwater with DU and
suppression of SNF dissolution, (2) maintenance of chemically reducing
conditions in the WP that minimize SNF solubility by sacrificial oxidation
of DU from the +4 valence state, and (3) evolution of DU to lower-density
hydrated uranium silicates. The fill expansion minimizes water flow in the
degraded WP. The DU also isotopically exchanges with SNF uranium as the SNF
degrades to reduce long-term nuclear-criticality concerns.
Study on the Formation of Lanthanum Phosphate for Highly
Efficient Incorporation of Actinides from Radioactive Liquid
Waste
Authors: Sasoh Michitaka , Miyamoto Shinya , Toyohara Masumitu , Wada Mikio
Published: September 2012

We have studied a co-precipitation method to remove actinides from
radioactive liquid waste. Lanthanum phosphate was selected as a
co-precipitation material because actinides and lanthanides are among the
homology series in the periodic table. In this report, the conditions, under
which the lanthanum compound is precipitated was looked at, as well as, ways
to widen the pH range in which the precipitation occurrs in sodium nitrate
solution. The properties of the resulting precipitates were also
investigated. Further, we investigated the incorporation ratio of amerícium
(decontamination factor), one of the actinides, into the precipitate from
sodium nitrate solution. Lanthanum phosphate was found to be more effective
compared to ferric compounds as the co-precipitation material to remove
amerícium. The incorporation conditions for strontium using lanthanum
phosphate were also investigated. The removal of strontium, by this method,
was more effective when the pH is above 7.0.
Formation of Paramagnetic Defects in Oxide Glasses During the
Bombardment of Their Surface with Charged Particles
Authors: Bogomolova L. D. , Stef Ano Vsky S. V. , Teplyakov Y. G. , Dmitriev S. A.
Published: September 2012

Heavy-ion bombardment of a glass surface is a conventional laboratory
technique for producing damage of interest for radioactive waste
encapsulation. At energy of order 100 keV such a bombardment simulates the
damage produced by α-recoil nuclei and fission fragments resulting from the
nuclear decay. The damage region is 100–500 nm depending on conditions of
the bombardment.

In the present work some results of EPR study of point defects formed in
silicate, borate, borosilicate, phosphate and other oxide glasses irradiated
with different charge particles (C, N, O, Ar. Mn, Cu, Pb) at energy E=150
keV and large total fluence of ions (up to 1017 cm-2)
are reported. Electron paramagnetic resonance (EPR) is a very sensitive
technique which gives an information on the structure of point defects and
their content. It is shown that in some cases (for example, in borate
glasses) the oxygen hole centers similar to ones observed in γ-irradiated
glasses are formed after ion bombardment. However, in the majority of cases
new defects which are not typical of γ-irradiated oxide glasses were found
They were large molecular oxygen ions (O2-O3-O4-) located in the cavities formed under ion bombardment in the
near surface layer of glass. It should be noted that the relative content of
these defects is of the order of several tens per 1000 incident ions. This
content decreases with increasing fluence and atomic mass of incident ions.
It indicates indirectly that point defects are clustered when the damage of
the near surface layer becomes strong. The formation of gaseous oxygen is
possible in cavities of the damage surface layer.

It was found that some elements (for example C, N and transition metals)
form chemical compounds with oxygen. The migration of alkali ions promotes
the formation of such compounds since the chemical compounds were detected
by means EPR in glasses rich in alkali oxides.
Numerical Predictions of Dry Oxidation of Iron and Low-Carbon
Steel at Moderately Elevated Temperatures
Authors: Henshall G. A.
Published: September 2012

Wrought and cast low-carbon steel are candidate materials for the thick
(e.g. 10 cm) outer barrier of nuclear waste packages being considered for
use in the potential geological repository at Yucca Mountain. Dry oxidation
is one potential degradation mode for these materials at the moderately
elevated temperatures expected at the container surface, e.g. 323–533 K
(50–260 °C). Therefore, numerical predictions of dry oxidation damage have
been made based on experimental data for iron and low-carbon steel and the
theory of parabolic oxidation. A numerical approach employing the Forward
Euler method has been implemented to integrate the parabolic rate law for
arbitrary, complex temperature histories. Assuming growth of a defect-free,
adherent oxide, the surface penetration of a low-carbon steel barrier
following 5000 years of exposure to a severe, but repository-relevant,
temperature history is predicted to be only about 0.127 mm, less than 0.13%
of the expected container thickness of 10 cm. Allowing the oxide to spall
upon reaching a critical thickness increases the predicted metal penetration
values, but degradation is still computed to be negligible. Based on these
physically-based model calculations, dry oxidation is not expected to
significantly degrade the performance of thick, corrosion allowance barriers
constructed of low-carbon steel.
Development of Mathematical Models for Long-Term Prediction of
Corrosion Behaviour of Carbon Steel Overpacks for Radioactive Waste
Disposal
Authors: Hoch A. R. , Honda A. , Porter F. M. , Sharland S. M. , Taniguchi N.
Published: September 2012

Mathematical models to enable long-term prediction of the corrosion
behaviour of carbon steel overpacks for radioactive waste have been
developed. An existing model of the growth of pits, implemented in the CAMLE
software, has been extended and used to investigate the sensitivity of the
predictions to input parameters, including cathodic reaction kinetics and
the relative position of the anode and cathode. Predictions have also been
made of the aeration period of the repository, during which localised
corrosion is possible.
Spectroscopic Investigation of Actinide Speciation in
Concentrated Chloride Solution
Authors: Runde W. , Neu M. P. , Conradson S. D. , Clark D. L. , Palmer P. D. , Reilly S. D. , Scott B. L. , Tait C. D.
Published: September 2012

Actinide solubilities in highly concentrated chloride solutions are about
one order of magnitude higher than in similar inert electrolyte
(NaClO4) solutions. This increased solubility is due to
interactions between actinide and chloride ions. Contradictory results exist
regarding the interaction mechanism between actinide and chloride ions.
Specifically, both inner-sphere complex formation and ion pair association
have been implicated in the interpretation of spectrophotometric and
extraction data. To address this controversy, we investigated the
interaction between actinide ions in the (III), (IV), (V) and (VI) oxidation
states and chloride ions using a multi-method approach. Spectroscopie
techniques (TRLFS, Raman, UV-Vis absorption, EXAFS) were used to distinguish
between changes in the inner coordination sphere of the actinide ion and
effects of ion pairing. X-ray absorption spectroscopy and single crystal
X-ray diffraction were used to determine structural details of the actinide
chloro complexes formed in solution and solid states.
Retrograde Solubilities of Source Term Phases
Authors: Murphy William M.
Published: September 2012

Natural analog, experimental, and thermodynamic studies indicate that the
properties of secondary uranyl minerals are likely to control the source
term for U and other radioéléments incorporated in these phases in the
proposed Yucca Mountain repository. Thermodynamic calculations using data
from the 1992 NEA data base indicate an increase in the equilibrium constant
for schoepite dissolution from 103.1 to 104.8 with
decreasing temperature from 100° to 25°C, i.e., retrograde solubility.
Enthalpies for mineral transformation reactions that consume protons and
release cations are typically negative, suggesting that solubilities of
other uranyl phases such as uranophane increase more than that of schoepite
with decreasing temperature. Solubilities of mineral phases associated with
spent nuclear fuel will be initially relatively low under the elevated
repository temperature regime. As the temperature of the repository
decreases due to radioactive decay and heat dissipation, source term mineral
solubilities increase. The rate of release of U and other species is
controlled by a series of processes: transport of oxidants and flux of
water; oxidative dissolution of spent fuel; uranyl mineral precipitation;
uranyl mineral dissolution or transformation; and radionuclide transport.
Decreasing diffusion and reaction rates and increasing uranyl mineral
solubilities with decreasing temperature may lead to a change with time from
solubility to transport or reaction rate as a source term controlling
mechanism. Preservation of large quantities of uranyl minerals formed by
oxidation of uraninite and radiometrie ages of secondary uranophane at the
natural analog site at Peña Blanca indicate that oxidative alteration of
uraninite was fast relative to transport of U away from the deposit. The
successive formation of schoepite and uranophane in natural settings where
uraninite has been oxidized may represent a paragenesis reflecting
increasing temperature or increasing incorporation of environmental
components. In contrast, diminishing temperature conditions in a repository
source area could lead to the reverse sequence of mineral formation.
The Effects of the Chemical and Radiolytic Degradation of
Asphalt on Plutonium Solubility
Authors: Greenfield B. F. , Ito M. , Spindler M. W. , Williams S. J. , Yui M.
Published: September 2012

The influence of alkaline degradation or radiolytic degradation of asphalt
on plutonium solubility has been investigated. Asphalt has been contacted
with water, sodium hydroxide solution or concrete leachate at 80°C for
periods of up to approximately 2 years. Sodium nitrate was also present in
some of the experiments. Plutonium solubilities were measured at pH 12 in
the leachates and found to be less than 10-8 mol dm-3
for most degradations. Relatively low levels of Total Organic Carbon were
measured in the leachates. Alpha radiolysis of asphalt in the presence of
concrete and water has also been studied. Samples of asphalt were
encapsulated in concrete after coating with the
238PuO2, crushed and leached at room temperature.
The solubility of plutonium was measured in samples of the leachates after
approximately 90 days and 180 days had elapsed. The results showed that the
solubility of plutonium in the α-radiolysis leachates remained low and was
in the range 2 × 10-11 to 8 × 10-9 mol
dm-3. A consideration of these results, and data published
elsewhere, suggests that chemical and radiolytic attack on asphalt or
bitumen under anaerobic, alkaline conditions typical of a deep cementitious
repository is unlikely to generate complexants for plutonium which are
effective at high pH. Any enhancement of plutonium solubility is likely to
be less significant than that arising from the degradation of some other
organic materials.
Characterization of Th Carbonate Solutions Using XAS and
Implications for Thermodynamic Modeling
Authors: Hess N. J. , Felmy A. R. , Rai D. , Conradson S. D.
Published: September 2012

The chemical behavior of actinide elements in tank solutions, in soil, and
in groundwater is dependent upon the chemical species that form when aqueous
solutions come in contact with the actinide compounds. In particular the
chemical speciation of the reduced actinide oxidation states (III and IV)
are important, for example, to DOE waste tank processing and, more
generally, to nuclear waste disposal issues. Predicting the solubility of
the actinides in these solutions requires identification of the strong
aqueous complexes, such as carbonates and organic chelating agents, that can
form in aqueous solution.

Previous speciation work has often relied on indirect techniques such as
potentiometric titrations or solubility measurements. Recent XAS experiments
determine directly the speciation of the Th carbonato species of seven
solutions under a range of carbonate concentrations and pH conditions. The
presence of the pentacarbonato complex is confirmed and the complex's
stability at low carbonate concentrations is determined. These experimental
results support a proposed thermodynamic model that describes the solubility
of Th(IV) hydrous oxide in the aqueous Na+-HCO3--CO32--OH--ClO4--H2O system extending to high concentrations at 25°C.
This model is relatively simple in that only two aqueous species are
included Th(OH)3CO3- and Th(CO3)56-.
The Effect of Mineral Variability on the Solubility of Some
Actinides: an Uncertainty Analysis
Authors: Ekberg Christian , Emrén Allan T. , Samuelsson Anders
Published: September 2012

The use of computer simulations in the performance assessment for a
repository for spent nuclear fuel, are in many cases the only method to get
information on how the rock-repository system will work. One important
factor is the solubility of the elements released if the repository is
breached. This solubility may be determined experimentally or simulated.
Ifit is simulated, several factors such as thermodynamical uncertainties
will affect the reliability of the results. If these uncertainties are
assumed to be small, the composition of the water used in the calculations
may play a major part in the uncertainties in solubility. The water
composition, in tum, is either determined experimentally or calculated
through water-rock interactions. Thus, if the mineral composition of the
rock is known, it is possible to foresee the water composition. However, in
most cases a determination of the rock composition is made from drilling
cores and is thus quite uncertain. Therefore, if solubility calculations are
to be based on water properties calculated from rock-water interactions
another uncertainty is introduced. This paper is focused on uncertainty and
sensitivity analysis of rock-water interaction simulations and the
uncertainties thus obtained are propagated through a program making
uncertainty and sensitivity analysis of the solubility calculations. In both
cases the latin hypercube sampling technique have been used. The results
show that the solubilities are in most cases log normal distributed while
the different elements in the simulated groundwater in some cases diverge
significantly from such a distribution. The numerical results are comforting
in that the uncertainty intervals of the solubilities are rather small, i.e.
up to 30%.
The Use of Laboratory Adsorption Data and Models to Predict Radionuclide Releases from a Geological Repository: a Brief History
Authors: Langmuir Donald
Published: September 2012

Radionuclide (RN) adsorption has long been recognized as important to assure the isolation of nuclear wastes in a geological repository [1]. Laboratory measured RN adsorption data have generally been expressed as distribution coefficient (Kd) values or adsorption isotherms. The proper application of these models is to site conditions nearly identical to those used in the laboratory adsorption experiments. This has required that multiple Kd's and isotherms be determined in a wide range of experiments designed to bracket expected repository conditions.

The surface complexation (SC) adsorption models were introduced in the late 1970's. The best known of these models incorporate electrical double layer (EDL) theory [2]. Their use requires that the water chemistry and surface properties of adsorbing rocks and minerals be fully characterized. Adsorption is then studied as reactions involving specific aqueous RN species (often complexes) and specific surface sites. Because the SC models are relatively mechanistic, they may allow extrapolation of adsorption results to repository conditions that lie outside the limited experimental range used to parameterize a given model. Turner [3] has shown that the diffuse layer model (the simplest SC model) fits a wide range of RN adsorption data as well as the more complex models. Others have suggested ways to generalize and estimate SC model parameters for a variety of minerals, rocks and engineered materials (cf. [4,5,6,7,8,9,10,11,12]. Degueldre and Werlni [12] and Degueldre et al. [13] have proposed a simplified SC model for RN adsorption that avoids EDL theory, in which the adsorption of RN species is estimated from linear free energy relationships.

It is appropriate to ask how accurately RN adsorption behavior must be known or understood for total system performance analysis (TSPA). In most geological settings now being considered for repository development globally, it may suffice to select bounding Kdvalues for the different rock types (cf. [14,15]). Use of the SC models to describe RN adsorption can provide us with increased confidence that minimum Kd's and the distribution of Kdvalues we might propose for TSPA are in fact conservative.
Influence of Carbonate Ions on Europium Sorption onto Iron-Oxides
Authors: Fujita T. , Tsukamoto M.
Published: September 2012

The sorption behavior of europium onto two iron-oxides, goethite and magnetite, focused on the effect of carbonate species concentration was investigated in 0.01 M NaClO4solution by the batch method over a pH range of 4 to 11. It was found that sorption of europium was enhanced in the higher concentration solution of carbonate species for both materials. The zeta-potential of goethite in the solutions including sodium bicarbonate was also measured to estimate the effect of sorbed carbonate species. Our experimental results and model calculation suggest that the production of surface carboxylic group FeOCOO by the sorption of carbonate species lowered the zeta-potential and enhanced europium sorption.
Liquidus Temperature Model for Hanford High-Level Waste Glasses
with High Concentrations of Zirconia
Authors: Crum J. V. , Schweiger M. J. , Hrma P. , Vienna J. D.
Published: September 2012

A study was conducted on glasses based on a simulated transuranic waste with
high concentrations of ZrO2and Bi2O3 to
determine the compositional dependence of primary crystalline phases and
liquidus temperature (TL). Starting from a baseline composition, glasses were formulated by
changing mass fractions of Al2O3,
B2O3, Bi2O3, CeO2,
Li2O, Na2O, P2O5,
SiO2, and ZrO2, one at a time, while keeping the
remaining components in the same relative proportions as in the baseline
glass. Liquidus temperature was measured by heat treating glass samples for
24 h in a uniform temperature furnace. The primary crystalline phase in the
baseline glass and the majority of the glasses was zircon
(ZrSiO4). A change in the concentration of certain components
(Al2O3, ZrO2, Li2O,
B2O3 and SiO2) changed the primary phase
to baddeleyite (ZrO2), while cerium oxide (CeO2)
precipitated from glasses with more than 3 wt% CeO2. Zircon
TL was strongly increased by Al2O3,
Zrb2 and CeO2, and slightly by
P2O5 and SiO2; decreased strongly by
Li2O and Na2O and moderately by
B2O3. A first-order model was constructed for
TL as a function of composition for zircon primary crystalline phase
glass.
Uranyl Sorption Onto Alumina
Authors: Jacobsson Anna-Maria M. , Rundberg Robert S.
Published: September 2012

The mechanism for the adsorption of uranyl onto alumina from aqueous
solution was studied experimentally and the data were modeled using a triple
layer surface complexation model. The experiments were carried out at low
uranium concentrations (9×10-11 - 5×10-8M) in a
CO2 free environment at varying electrolyte concentrations
(0.01 – 1 M) and pH (4.5 – 12). The first and second acid dissociation
constants, pKal and pKa2, of the alumina surface were
determined from potentiometric titrations to be 7.2 ± 0.6 and 11.2 ± 0.4,
respectively. The adsorption of uranium was found to be independent of the
electrolyte concentration. We therefore conclude that the uranium binds as
an inner sphere complex. The results were modeled using the code FITEQL. Two
reactions of uranium with the surface were needed to fit the data, one
forming a uranyl complex with a single surface hydroxyl and the other
forming a bridged or bidentate complex reacting with two surface hydroxyls
of the alumina. There was no evidence from these experiments of site
heterogeneity. The constants used for the reactions were based in part on
predictions made utilizing the Hard Soft Acid Base, HSAB, theory, relating
the surface complexation constants to the hydrolysis of the sorbing metal
ion and the acid dissociation constants of the mineral oxide surface.
Adsorption of Sn(IV) on Goethite in 0.01M NaCl Solution at
Ambient Temperature
Authors: Amaya Takayuki , Suzuki Kazunori , Oda Chie , Yoshdcawa Hideki , Yui Mikazu
Published: September 2012

The adsorption of soluble Sn(TV) species on goethite (FeOOH), a naturally
occurring mineral, was investigated in 0.01 M NaCl solution at ambient
temperature. The sequential spike method was used to obtain the distribution
coefficient (Kd). Kd values of higher than 1×107 ml/g at neutral
pH, and 3×105 ml/g at pH 11 were obtained. It is observed that Kd
values decrease with increasing pH. In the desorption experiment, the
sequential extraction method was applied. The extractants such as 0.1 M
NaOH, Tamm's acid oxalate and Coffin's reagent were used. The result
indicates that the adsorption consists of both reversible and irreversible
reactions, and that the most of fractions in adsorption of Sn on goethite
correspond to the irreversible adsorption or a stable fixation. Moreover, it
is suggested that the reversible adsorption converts to the irreversible
adsorption with time.
Sorption Processes of UO22+Ions onto Zirconium and Thorium Phosphate Surfaces: Thermodynamical and Structural Aspects
Authors: Drol R. , Simoni E. , Cavcllcc R. , Pradin V. , Dcnauwcr Ch. , Ehrhardl J. J.
Published: September 2012

Most of the studies on sorption processes take into account only a ihcrmodynamic approach, and only lew studies have paid attention to ihc structural aspect (i.e. structure of ihc postulated surface complex). Experiments were conducted to study UO22+sorption onto zirconium and thorium phosphate surfaces. Moreover, in order to identify the nature of the sorption sites and the structure of the surface complex, various spectroscopie techniques such as EXAFS, XPS and laser induced fluorescence have been used. Two different types of sorption sites on the thorium phosphate but only one on the zirconium phosphate have been detected.
Investigation of the Influence of Heterogeneous Porosity on
Matrix Diffusion: a Novel Approach Using Adaptive Tree-Multigrid Technique
and Real Porosity Data
Authors: Simbierowicz P. , Olin M.
Published: September 2012

Last year we developed a two-dimensional deterministic heterogeneous matrix
diffusion model, which is capable of utilising porosity information
originating from real drill-core samples. The results of numerical
infiltration experiment we had performed with the model displayed
substantial spatial variations in the penetration depth. Because it is
practically impossible to verify experimentally those two-dimensional
penetration profiles we had computed, this time we decided to try modelling
of measured leaching curves. Unfortunately we have not succeeded in
acquiring such curves for the exact same samples, which we have used in
numerical leaching experiments. Nevertheless it can be seen, that the shape
of leaching curves computed with the heterogeneous model is clearly closer
to the shape of measured curves, than the shape of curves provided by the
standard model. These differences can be utilised as a basis for an
approximate numerical method of assessing the geometric factor, which has
traditionally been a purely empirical parameter.

The results of the new numerical experiments agree with our older results
from last year: the heterogeneity of the rock matrix has highly significant
impact on the diffusion. However, when interpreting the results, one must
not neglect numerous limitations of the model, and hence, one should not
attempt to overgeneralise the conclusions.
A Note on Radionuclide Transport by Gas Bubbles
Authors: Neretnieks Ivars , Ernstson Märta-Lena
Published: September 2012

In a repository for spent nuclear fuel, gas generated by corrosion of the
iron in the canister may form small bubbles that will escape and rise to the
ground surface. Colloidal particles may attach to the surface of the bubbles
and be carried by them. If the colloids are supplied by the montmorillonite
clay of the buffer material surrounding the canister, the clay can be
carried away. Nuclides sorbed in the clay can be carried with the bubbles.
We have estimated the carrying capacity of the gas of the clay particles and
the escape rate of nuclides carried by the gas bubbles. The latter is also
compared to the escape rate by the conventional escape mechanisms from the
near field. We have further estimated the detachment of the nuclides from
the clay and their sorption onto the fracture surfaces of the rock as well
as their uptake by diffusion into the rock matrix along the bubble transport
paths.

The present paper is speculative and uses some hypothetical assumptions.
Although the processes that are modelled are known to exist there is not
enough known of several of them to quantify them accurately. The carrying
capacity of the gas used in the calculations is an upper bound and probably
very much exaggerated. Even so, the consequences are minor for the release
of radionuclides.
Waste Vitrification: Prediction of Acceptable Compositions in a
Lime-Soda-Silica Glass-Forming System
Authors: Gilliam T M. , Jantzen T. M.
Published: September 2012

A model is presented based upon calculated bridging oxygens which allows the
prediction of the region of acceptable glass compositions for a
lime-soda-silica glass-forming system containing mixed waste. The model can
be used to guide glass formulation studies (e.g., treatability studies) or
assess the applicability of vitrification to candidate waste streams.
Monte Carlo Simulation of Radioactive Contaminant Transport in
Fractured Geologic Media: Disorder and Long-Range
Correlations
Authors: Mukhopadhyay Sumit , Cushman John H.
Published: September 2012

The geologic media near Yucca mountain site consist of fractured welded
tuffs along with less fractured unwelded tuff. Numerical simulation of flow
and transport in such media poses a number of challenging problems, due
mainly to the heterogeneities and disorder in the media. In addition,
because of different dominant transport mechanisms in different regions of
the media, investigations need to be carried out at different time-scales.
“Time-marching” will pose a considerable problem in analyzing such
multi-scale transient problems. We develop a field-scale network model of
fractures and study transport of radionuclides through geologic media as a
function of disorder and correlated fracture-permeabilities.
Diffusion of Technetium in Compacted Bentonites in the Reducing Condition with Corrosion Products of Iron
Authors: Kuroda Yuji , Idemitsu K. , Furuya H. , Inagaki Y. , Arima T.
Published: September 2012

In the vicinity of a high-level waste repository, corrosion of carbon steel overpacks will create a reducing environment. Reducing conditions are expected to retard the migration of redox-sensitive radionuclides such as technetium.

The apparent diffusion coefficients of technetium were measured in compacted bentonites (Kunigel VI® and Kunipia F®, JAPAN) in contact with carbon steel and its corrosion products under reducing conditions or without carbon steel under oxidizing conditions for comparison. The apparent diffusion coefficients measured were 10-12to 10-13m2/s under oxidizing conditions and 10-12to 10-13m2/s under reducing conditions. There were significant effects of redox condition, dry density (0.2 to 2.3 g / cm3) and montmorillonite content (50% for Kunigel VI or 100% for Kunipia F) on the apparent diffusion coefficients. Montmorillonite density could be a good index to explain density dependence of the diffusion coefficients under both reducing and oxidizing conditions.
Detailed Description of a Long-Term Low-Level Waste Degradation
Experiment
Authors: Caron F. , Torok J. , Haas M. K. , Manni G.
Published: September 2012

This work gives a detailed description of the important aspects of a
long-term Low-Level Radioactive Waste (LLRW) degradation experiment,
performed at Chalk River Laboratories (CRL). This experiment utilized actual
LLRW. The wastes consist of unconditioned compacted refuse (paper, mop
heads, paper towels, used clothing, etc), which represents the bulk of the
waste volume intended for near-surface disposal at CRL. Waste material was
collected and compacted to make a total of 11 bales for this experiment.
Each bale was then placed and sealed in separate steel containers which were
connected to sampling lines. After a dry monitoring period, water was added
to promote leaching and decomposition of the wastes. The leachate sampled
had a composition similar to landfill leachates. Some applications of this
experiment, used to support the safety case of near-surface disposal, are
briefly discussed in this paper, e.g., the production of colloidal material,
the nature and role of dissolved organics of microbial origin, etc.
Enumeration of Microbial Populations in Radioactive Environments by Epifluorescence Microscopy
Authors: Pansoy-Hjelvik M. E. , Strietelmeier B. A. , Paffett M. T. , Kitten S. M. , Leonard P. A. , Dunn M. , Gillow J. B. , Dodge C. J. , Villarreal R. , Triay I. R. , Francis A. J.
Published: September 2012

Epifluorescence microscopy was utilized to enumerate halophilic bacterial populations in two studies involving inoculated, actual radioactive waste/brine mixtures and pure brine solutions. The studies include an initial set of experiments designed to elucidate potential transformations of actinide-containing wastes under salt-repository conditions, including microbially mediated changes.

The first study included periodic enumeration of bacterial populations of a mixed inoculum initially added to a collection of test containers. The contents of the test containers are the different types of actual radioactive waste that could potentially be stored in nuclear waste repositories in a salt environment. The transuranic waste was generated from materials used in actinide laboratory research. The results show that cell numbers decreased with time. Sorption of the bacteria to solid surfaces in the test system is discussed as a possible mechanism for the decrease in cell numbers.

The second study was designed to determine radiological and/or chemical effects of239Pu,243Am,237Np,232Th and238U on the growth of pure and mixed anaerobic, denitrifying bacterial cultures in brine media. Pu, Am, and Np isotopes at concentrations of ≤1×10–5M, ≤5×10–6M and ≤5×10-4M respectively, and Th and U isotopes at concentrations of ≤4×10-3M were tested in these media. The results indicate that high actinide concentrations affected both the bacterial growth rate and morphology. However, relatively minor effects from Am were observed at all tested concentrations with the pure culture.