bluuebunny/crossref_metadata_2025_split huggingface dataset:
Source-Term Analysis for Hanford Low-Activity Tank Waste Using
the Reaction-Transport Code AREST-CT
Authors: Chen Y. , McGrail B. P. , Engel D. W. 
Published: September 2012

ABSTRACTA general, integrated performance assessment code, AREST-CT, was used to
analyze the influence of various factors on the release rates of
radionuclides from a proposed facility for disposal of low-activity tank
wastes. The code couples various process models together based on the
framework of reaction-transport theory. The disposal facility was modeled as
a 1-D column surrounded by soil. A borosilicate waste glass, LD6–5412 was
the waste form considered in the analysis. Included in the simulations were
38 aqueous species, 14 minerals, 21 equilibrium reactions, and 16 kinetic
reactions. Dissolution rate of the glass and the release rates of Te, Pu, U,
Np, I, Se under different conditions were calculated for 50,000 years. The
simulations revealed that 1) open exchange between the atmosphere and
pore-water within the vault significantly improves the performance; 2) an
ion-exchange reaction between the glass and aqueous phase increases the
release rates significantly; and 3) at the hydrogeologie conditions under
consideration, variation of the pore-water velocity has little effect on the
release rate of radionuclides. These results provide a scientific basis for
formulation of waste forms and engineering design of the disposal facility.
Reaction-transport modeling can provide information on the long-term
performance of disposal systems that are not obtainable from laboratory
experiments alone or by conventional decoupled process models.
A Mathematical Model for Advective Release from Perforated Waste
Package Container Under Dripping Water
Authors: Lee Joon H. 
Published: September 2012

ABSTRACTA defense-in-depth engineered barrier system (EBS) is employed in the
current design concept for the potential high-level nuclear waste repository
at Yucca Mountain, Nevada, USA. Simplifying the geometry of the cylindrical
waste container into the equivalent spherical configuration, and
incorporating detailed analysis of the mechanics of water flow around the
waste container surface, a mathematical model is developed for advective
release from a “failed” (or perforated) waste container under dripping
water. It is shown that the advective release rates are controlled by
diffusion through the perforations in the waste container, and affected
insignificantly by the dripping flow rate for the flow
rate range considered. The release rates depend strongly on the number of
perforations (or pit penetrations) in the waste container. The insensitivity
of the release rate to the dripping flow rate is explained by the fact that
radionuclide is released from the container surface to the boundary layer of
the water film which contacts the container surface and is relatively
stagnant. Also, since a laminar flow around the waste container surface is
assumed in the model development, radionuclides transport across the water
layers in the film by diffusion only. Additionally, the insignificant effect
of the flow rate is contributed by the “short” penetration depth of
radionuclide into the water film assumed in the model development. The
analyses show that the number of perforations, the size of perforation, the
container wall thickness, and the geometry (i.e., radius) of the waste
container are important parameters that control the advective release rate.
It is emphasized that the “failed” (or perforated) waste package container
can still perform as a potentially important barrier to radionuclide
release.
Bounding Analysis for Solubility
Authors: Ohi T. , Nakajima K. 
Published: September 2012

ABSTRACTIn performance assessment of geological disposal systems for High Level
Radioactive Waste (HLW), the change of environment over the long-term must
be considered. Therefore, it is necessary to consider a wide range of
parameters concerned with radionuclides migration, especially the dependence
of solubility on geochemical environment. In this study, assuming that the
release rate of the nuclides from buffer material is limited by inventory
ultimately, the relationship between the initial inventory and the
solubility that produces a solubility-invariant maximum release rate from
the buffer is examined by using a simple steady-state analytical solution
without decay. As the result, the threshold of “effective” solubility in the
performance assessment of the geological disposal systems for HLW is
obtained as a function of initial inventory, distribution coefficient (Kd),
diffusion coefficient, and thickness, porosity and density of the buffer.
Also, the threshold of “effective” steady dissolution rate corresponding to
the threshold of “effective” solubility is obtained.
Americium/Curium Extraction from a Lanthanide Borosilicate
Glass
Authors: Rudisill T. S. , Pareizs J. M. , Ramsey W. G. 
Published: September 2012

ABSTRACTA solution containing kilogram quantities of highly radioactive isotopes of
amerícium and curium (Am/Cm) and lanthanide fission products is currently
stored in a process tank at the Department of Energy's Savannah River Site
(SRS). This tank and its vital support systems are old, subject to
deterioration, and prone to possible leakage. For this reason, a program has
been initiated to stabilize this material as a lanthanide borosilicate (LBS)
glass.1 The Am/Cm has commercial value and is desired for use
by the heavy isotope programs at the Oak Ridge National Laboratory
(ORNL).A recovery flowsheet was demonstrated using a curium-containing glass to
extract the Am/Cm from the glass matrix. The procedure involved grinding the
glass to less than 200 mesh and dissolving in concentrated nitric acid at
110°C. Under these conditions, the dissolution was essentially 100% after 2
hours except for the insoluble silicon. Using a nonradioactive surrogate,
the expected glass dissolution rate during Am/Cm recovery was bracketed by
using both static and agitated conditions. The measured rates, 0.0082 and
0.040 g/hrcm2, were used to develop a predictive model for the
time required to dissolve a spherical glass particle in terms of the glass
density, particle size, and measured rate. The calculated dissolution time
was in agreement with the experimental observation that the curium glass
dissolution was complete in less than 2 hr.
The Wipp Pre-Disposal Phase Experimental Program
Authors: Matthews M. L. 
Published: September 2012

ABSTRACTThe WIPP is a transuranic waste disposal project of the U.S. Department of
Energy and administered by the Carlsbad Area Office. The WIPP facility is a
mined repository located near Carlsbad, New Mexico. It was developed and
constructed 658 meters below the surface to demonstrate the safe, permanent
isolation of transuranic radioactive wastes in a bedded salt formation.
After the site selection phase and construction phase were completed, DOE
embarked upon a pre-disposal experimental program phase. There are four
major WIPP experimental program areas: shaft seals and rock mechanics;
disposal room interactions; fluid flow and transport; and direct releases
due to human intrusion.The Pre-Disposal Phase Experimental Program provided the necessary
information to support the WIPP performance assessment process. The results
of the program have been instrumental in demonstrating that the WIPP meets
the regulatory requirements.
The Hyrkkölä Native Copper Mineralization: a Natural Analogue
for Copper Canisters
Authors: Perea Nuria Marcos 
Published: September 2012

ABSTRACTThe Hyrkkölä U-Cu mineralization is located in south-western Finland, near
the Palmottu analog site, in crystalline, metamorphic bedrock. The age of
the mineralization is estimated to be between 1.8 and 1.7 Ga. The existence
of native copper and copper sulfides in open fractures in the near-surface
zone allows us to study the native copper corrosion process in conditions
analogous to a nuclear fuel waste repository.From the study of mineral assemblages or paragenesis, it appears that the
formation of copper sulfide (djurleite, Cu1.934S) after native
copper (Cu°) under anoxic (reducing) conditions is enhanced by the
availability of dissolved hydrogen sulfide (HS) in the groundwater
circulating in open fractures in the near-surface zone. The minimum
concentration of HS in the groundwater is estimated to be of the order of
10-5 M (∼ 10-4 g/1) and the minimum pH value not
lower than about 7.8 as indicated by the presence of calcite crystals in the
same fracture.The present study is the first one performed on occurrences of native copper
in reducing, neutral to slightly alkaline groundwaters. Thus, the data
obtained is of most relevance in improving models of anoxic corrosion of
copper canisters.
Mechanisms for the Formation of a Perched Water Zone in
Fractured Tuff: a Natural Analogue Study
Authors: Woodhouse E. G. , Bassett R. L. 
Published: September 2012

ABSTRACTPerched water zones have been identified in the fractured, welded tuff in
the semi-arid to arid environments of Yucca Mountain, Nevada and near
Superior, Arizona. An understanding of the formation of such zones is
necessary in order to predict where future perched water might form at Yucca
Mountain, the proposed site of a high-level nuclear waste repository. The
formation or growth of a perched zone near a repository is one of the
factors to be considered in the risk assessment of the Yucca Mountain
site.The Apache Leap Research Site near Superior, Arizona is a natural analog to
the Yucca Mountain site in terms of geology, hydrology, and climate.
Information used to study possible mechanisms for the formation of the
perched zone included data regarding isotopie and geochemical properties of
the waters in and above the perched water zone; measured hydrologie
parameters of the perched zone; geophysical and measured parameters of the
tuff; megascopic and microscopic observations of the tuff, including
mineralogical, alteration, and structural features; and the lateral and
vertical extent of perched water in the region.Aquifer test, geophysical, geochemical, and radioisotopic data show that
fractures are the means by which water is recharging the perched zone. The
reduced hydraulic conductivity of the formation in the perched zone appears
to result from both a severe reduction in matrix porosity and permeability
caused by welding, devitrification, and vapor phase crystallization; and by
an increase in fracture filling which restricts the pathways for flow.
Processing and Characterization of Glass Composites for ASH
Immobilization
Authors: Sobolev I. A. , Ojovan M. I. , Karlina O. K. , Stefanovsky S. V. , Galtsev V. E. 
Published: September 2012

ABSTRACTInduction melting by using a cold crucible is a suitable technology for the
immobilisation of ash residues after incineration of solid radioactive
waste. We investigated the possibility of using glass composites produced by
stirring the ash into meltedglass. Glass composites containing 15 -40 wt. %
of ash were obtained in both laboratory and bench scale units. Infrared
spectroscopy, electronic paramagnetic resonance, X-ray diffractometry, TEM
and SEM analyses were applied in order to characterise the structure of the
glass composites obtained. The glass composites consisted of a relatively
homogeneous glass matrix with embedded polycrystalline aggregates. The
fraction of aggregates increases when the fraction of ash rises. The
isothermal curing of composites at 1100 °C leads to dissolution of the ash
components into the melt as well as to their inclusion into the glass
structure, according to the analysis of the spectra obtained.
Thermodynamic Stabilities of U(VI) Minerals: Estimated and
Observed Relationships
Authors: Finch Robert J. 
Published: September 2012

ABSTRACTGibbs free energies of formation (ΔG°ƒ) for several
structurally related U(VI) minerals are estimated by summing the Gibbs
energy contributions from component oxides. The estimated ΔG°f values are used to construct activity-activity (stability)
diagrams, and the predicted stability fields are compared with observed
mineral occurrences and reaction pathways. With some exceptions, natural
occurrences agree well with the mineral stability fields estimated for the
systems Sio2-Cao-Uo3-UOH2O and
Co2-caO-UO3-H2O providing confidence in
the estimated thermodynamic values. Activity-activity diagrams are sensitive
to small differences in ΔG°f values, and mineral compositions must be known accurately,
including structurally bound H2O. The estimated ΔG°f values are not considered reliable for a few minerals for two
major reasons: (1) the structures of the minerals in question are not
closely similar to those used to estimate the ΔG°f* values of the component oxides, and/or (2) the minerals in
question are exceptionally fine grained, leading to large surface energies
that increase the effective mineral solubilities.
Uraninite-Water Interactions in an Oxidizing
Environment
Authors: Fayek M. , Kyser T. K. , Ewing R. C. , Miller M. L. 
Published: September 2012

ABSTRACTExceptionally low δ 18O values of primary uraninite and
pitchblende (i.e. -32 per mil to -19.5 per mil) from Proterozoic
unconformity-type uranium deposits in Saskatchewan, Canada, in conjunction
with theoretical uraninite-water oxygen isotope fractionation factors
suggest that primary uranium mineralization is not in oxygen isotopie
equilibrium with clays and silicates. The low δ 18O values have
been interpreted to have resulted from the recrystallization of primary
uranium mineralization in the presence of modern meteoric fluids having low
δ 18O values of ca. -18 per mil. The absence of apparent
alteration in many of the uraninite and pitchblende samples requires that
the uranium minerals exchange oxygen isotopes with fluids, with only minor
disturbances to their original chemical compositions and textures. However,
experiments on the interaction between water and natural uraninites, from
these deposits, and detailed electron micro-probe analyses of natural
uraninite and pitchblende indicate that, in the presence of water, old
uraninite rapidly alters to curite
(Pb2U5O174H2O). The hydration
of uraninite to curite releases uranium and calcium into solution and
becquerelite
(Ca(UO2)6O4(OH)6H2O)
is precipitated. In the presence of Si-saturated waters, uranium silicate
minerals, soddyite
((UO2)2(SiO4)2H2O) and
kasolite (Pb(UO2)SiO4H2O are precipitated
in addition to, curite and schoepite
((UO2)8O2(OH)12(H2O)12).
The mineral paragenesis observed in these experiments is similar to
sequences observed in oxidized zones in uranium deposits and
UO2-water experiments. Therefore, it is unlikely that natural
uraninite and pitchblende can simply exchange oxygen with an oxidizing fluid
without a concomitant change in phase chemistry or structure, nor will
oxidation of uraninite lead to the formation of U3O7.
as predicted by theoretical calculations used in natural analogue studies
for the disposal of high level nuclear waste.
Uraninite: a 2 Ga Spent Nuclear Fuel from the Natural Fission
Reactor at Bangombé in Gabon, West Africa
Authors: Jensen K. A. , Ewing R. C. , Gauthier-Lafaye F. 
Published: September 2012

ABSTRACTUraninites from the Bangombé natural fission reactor (RZB) and “normal”
uranium-ore occur as fine veins in the sandstone host-rock as well as
altered, broken, and slightly displaced grains in an illitic matrix, and in
nodules and veins of solid bitumen. Inclusions of galena, (Y,Gd)-rich
phosphates, a Pb-oxide and a Ti-oxide? were observed. Uraninites just below
RZB were partially altered to a uranyl-sulfate. Three generations of
uraninite were identified based on their PbO-contents of 8–11.06 wt%, 6 wt%
(the largest population), and a younger generation with 3 wt%. The high
Pb-uraninites may be the precursor to the low Pb-uraninites. Diffusional
loss of Pb is indicated by the presence of a Pb-oxide at the interface to
the uraninites. The behaviour of the metallic fission products, incompatible
with the uraninite structure, may mimic the behaviour of Pb in these
uraninites. The averaged impurity-content ranges from 4.29 to 6.89 wt%, and
consists mainly of SiO2, TiO2, ZrO2, FeO,
CaO, Al2O3 and P2O5. The
averaged content of Y2O3 and the Ln's is less than
0.78 wt% and there is a scattered positive correlation with
P2O5. The content of Y + Ln's is generally highest
in the uraninites from RZB. Uraninite hydration and the formation of
“uranopelite/zippeite” have caused complete loss of Y and the Ln's. These
elements seems also to be partially lost by weak phosphatian
coffinitization. The analytical results indicate that Y and the Ln's, which
are high yield fission products, may be released from uraninite during
alteration in the presence of P.
Mineralogical and Microtextural Characterization of “Gel-Zircon”
from the Manibay Uranium Mine, Kazakhstan
Authors: Helean K. B. , Burakov B. E. , Anderson E. B. , Strykanova E. E. , Ushakov S. V. , Ewing R. C. 
Published: September 2012

ABSTRACT“Gel-zircon”, an unusual Zr-silicate phase from the Manibay uranium mine,
northern Kazakhstan, was studied using X-ray diffraction (XRD), electron
microprobe energy dispersive X-ray spectroscopy (EDS) and high resolution
transmission electron microscopy (HRTEM). XRD results indicate that
gel-zircon is mostly amorphous and occurs with numerous impurity phases.
Microprobe EDS results indicate a UO2 content up to 9.14 wt. %.
HRTEM images revealed that the microtexture of gel-zircon consists of
nanocry stal lites of zircon, 2–10 nm in size, in a dominantly amorphous
matrix. Despite the U-Pb age of 420±25 my and the lack of significant
crystallinity, the gel-zircon is an apparently chemically durable phase.
Leaching of uranium ores which contain gel-zircon as the major U-bearing
phase is impossible using existing uranium plant technologies. The
alpha-decay dose, 2.64 displacements per atom (dpa), corresponding to the
age of gel-zircon is much higher than that (0.5 dpa) required to cause
metamictization of crystalline zircon. However, the morphology of gel-zircon
which occurs as veins up to 5 mm thick and tens of mm long does not indicate
initial crystallinity. Initially crystalline natural zircons often preserve
their crystal morphology after metamictization. This amorphous phase is
analogous to the highly damaged state characteristic of zircon proposed as a
waste form for the disposition of excess weapons plutonium.
Development and Testing of a Glass Waste Form for the
Immobilization of Plutonium
Authors: Chamberlain D. B. , Hanchar J. M. , Emery J. W. , Hoh J. C. , Wolf S. F. , Finch R. J. , Bates J. K. , Ellison A. J. G. , Dingwell D. B. 
Published: September 2012

ABSTRACTTwo alkali-tin-silicate (ATS) glasses have been prepared at Argonne National
Laboratory (ANL) as part of our ongoing research in radioactive waste glass
development. These glasses dissolved 5% and approximately 7% Pu. Early
corrosion test results indicate that Pu-bearing ATS glass is extremely
durable. The initial goal in this project concerned equally both the
solubility of Pu and the durability of the ATS glasses; however, our primary
emphasis has changed recently to maximizing the loading of Pu in the glass.
ATS-based glasses, using Th(VI) and Ce(III) as surrogates for Pu(IV), are
now being investigated to increase the solubility of Pu without
substantially sacrificing the durability of the current ATS formulations.
The solution data from various corrosion tests on the original Pu-containing
ATS glasses are also presented.
The Use of Laser Microprobe-ICP-MS in the Examination of Nuclear
Waste Glasses
Authors: Shuttleworth S. , Monteith J. E. 
Published: September 2012

ABSTRACTVitrification is the most developed method for the immobilisation of highly
radioactive waste obtained from reprocessing irradiated nuclear fuel. Highly
active wastes continue to be vitrified at Sellafield in the Waste
Vitrification Plant (WVP) and the product glass has previously been
subjected to rigorous characterisation in order to ensure waste form
integrity and stability during interim storage. However, the vitrified waste
may eventually be placed in an underground repository and so further studies
may be needed to confirm the chemical stability of vitrified wastes in
different disposal scenarios. One area of uncertainty is in knowing how the
alteration products and surface layers on a corroded glass effect the
dissolution process and the solution concentrations of long-lived
radionuclides. This study shows how the technique of laser microprobe
inductively coupled plasma mass spectrometry (LM-ICP-MS) can be used to
provide a direct chemical analysis of vitrified samples, both in their
as-cast state and after leach testing. LM-ICP-MS, solutions' nebulisation
ICP-MS (to provide chemical analyses of the leachants), light microscopy,
scanning electron microscopy and confocal laser microscopy were used to
examine the effect of leaching on the surfaces of vitrified samples. Data
were obtained for a number of simulated WVP-type product glasses that were
subjected to soxhlet leach testing for varying lengths of time. The
LM-ICP-MS data clearly show that the technique offers a rapid and relatively
accurate method for detecting and monitoring minute changes in the surface
layers of glass and in the build-up of alteration products. Hence, LM-ICP-MS
could be used to obtain valuable mechanistic information on the degradation
of glass by examining radionuclide behaviour in the surface layers of glass
as it corrodes over extended time periods.
Composition Effects on Viscosity and Chemical Durability of
Simulated Plutonium Residue Glasses
Authors: Bulkley S. A. , Vienna J. D. 
Published: September 2012

ABSTRACTVitrification is currently being explored as an option for plutonium (Pu)
scrap and residue stabilization because of its ability to convert
plutonium-bearing materials into a safe, durable, and accountable waste form
with reduced need for safeguarding. To develop this glass, the effect of
composition on key glass properties (viscosity and chemical durability) were
measured. A one-component-at-a-time change test matrix with 36 glasses was
designed. The viscosity is reported as the temperature at 5 Pas (T5), and
the durability as the normalized release of boron and sodium (Гb and ГNa)
from MCC-1 and PCT static leach tests and the negative logarithm of
fractional aluminum and lithium release (-ln[fA1] and
-ln[fLi]) from TCLP. The effects of
Al2O3, B2O3, CaO,
CeO2, Gd2O3, K2O,
Li2O, MgO, Na2O, PbO, SiO2, SnO,
TiO2, and ZrO2 were found to be linear for
T5 and non-linear for chemical durability. T5 is
increased most by the addition of SiO2 followed by
Al2O3, and decreased most by Li2O
followed by Na2O. MCC-1 ln[rB] and ln [rNa]
are increased most by the addition of Li2O followed by MgO and
decreased most by Al2O3 followed by ZrO2.
PCT In[rB] and In[rNa] are increased most by the
addition of MgO followed by B2O3 and decreased most by
Al2O3 followed by SiO2. The relative
effect of components on rB and rNa are similar for
both PCT and MCC-1. TCLP -ln[fAl] and ln[fLi are
increased most by the addition of SiO2 followed by SnO and
decreased most by Li2O.
Actinide Solubility in Lanthanide Borosilicate Glass for
Possible Immobilization and Disposition
Authors: Meaker T. F. , Peeler D. K. , Marra J. C. , Pareizs J. M. , Ramsey W. G. 
Published: September 2012

ABSTRACTImmobilization by vitrification is one potential disposition option for a
portion of the United States’ excess plutonium inventory. Research has been
performed to determine the glass forming region of a frit, plutonium and
rare earth system. The frit contains mainly oxides of aluminum, silicon and
boron; small amounts of ZrO2 and SrO are also included. The rare
earth elements provide a flux to the glass during processing. The rare
earths are also added as neutron absorbers (to prohibit criticality) in the
final vitreous product.This report will show the compositional region, using Th as a Pu surrogate,
that should be targeted in future studies for maximum Pu solubility in the
lanthanide borosilicate glass system. Durability data and process variables
will also be provided.