bluuebunny/crossref_metadata_2025 huggingface dataset:
The Wipp Pre-Disposal Phase Experimental Program
Authors: Matthews M. L.
Published: September 2012

The 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

The 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

Perched 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

Induction 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

Gibbs 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

Exceptionally 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

Uraninites 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

“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

Two 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

Vitrification 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

Vitrification 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

Immobilization 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.
The 4th Unit of the Chernobyl NPP: Its Present State and Problems Related to its Transformation into an Ecologically Safe System
Authors: Anderson E. B. , Borovoy A. A. , Pazukhin E. M.
Published: September 2012

The following problems related to the destroyed 4thUnit of the Chernobyl Nuclear Power Plant (NPP) are considered.

Monitoring of the unit's current state, including:

1)

Monitoring of the atmospheric release of radioactive elements in the form of aerosols. In this case the problem is the unqualified area of cracks and holes that have formed in the “Sarcophagus” (the “Shelter”) since 1986. In addition, the method used to evaluate the release, in which an uncontrolled escape of radioactive contamination is possible, raises objections;

2)

Monitoring of the aquatic medium both inside the “Shelter” and on the industrial site. At present, the quantity of water in the 4thUnit is about 3000 m, which in fact is low- and intermediate-level radioactive waste. This quantity is not constant and essentially varies depending on the season, amount of atmospheric precipitation, the intensity of dust suppression, etc. In this case the problem is that the paths of water migration inside the unit are not known, so that an uncontrolled penetration of water into the ground, via the damaged foundation slab or the unit's walls, is quite possible;

3)

Monitoring of the state of lava-like fuel-containing masses (LFCM) and the fuel. The problem is that the exact location of the fuel in the 4thUnit, the amount of fuel in major aggregates of LFCM, and the nuclear safety of these aggregates are all unknown to date.

The arrangements for the stabilization of the 4thUnit first of all include the reinforcement of the structural construction.

The transformation of the “Shelter” into an ecologically safe system is subject to many questions still unresolved. In particular:

Whether to create a common “Shelter-2” for the 3rand 4thUnits or to merely encase the destroyed 4thUnit in concrete (the project “Cube”);

Whether to peace a plant for conditioning and reprocessing of radioactive waste under the roof of the “Shelter-2”, or to build this plant separately;

Where to create a temporary repository for containers of radioactive waste from the ChNPP 4thUnit;

How to solve the problem of radioactive waste permanent storage in the territory of the Ukraine with proper observation of all nuclear and radiation safety requirements
The Behavior of Nuclear Fuel in First Days of the Chernobyl Accident
Authors: Burakov B. E. , Anderson E. B. , Shabalev S. I. , Strykanova E. E. , Ushakov S. V. , Trotabas M. , Blanc J-Y. , Winter P. , Duco J.
Published: September 2012

Various types of Chernobyl fuel containing masses named black “lava”, brown “lava”, porous “ceramic” and “hot” particles that formed during first days of the accident at the Chernobyl Nuclear Power Plant 4th Unit were studied by methods of optical and electron microscopy, microprobe and x-ray diffraction. Data about their chemical, phase and radionuclide composition are summarized. The products of interaction between fuel, zircaloy and concrete, produced under experiments in laboratory were examined for comparison with samples of Chernobyl “lava” and “hot” particles. The behavior of nuclear fuel in first days of the Chernobyl accident was a three-stage process. The first stage occurred before the moment of the Chernobyl explosion and was exceptionally short-lasting, perhaps, less than a few seconds. It was characterized by reaching a high temperature, ≥2600 °C, in the epicenter of accident and formation of a Zr-U-O melt in a local part of the core, which is estimated to be not more than 30% of whole core volume. The second stage lasted for about 6 days since the explosion, during which there was interaction between uranium products of the destroyed reactor: UOx, UOx with Zr, Zr-U-O, with the environment and silicate structural materials of the 4th Unit. The third stage, after 6 days involved the process of final formation of the radioactive silicate melt or Chernobyl “lava” at one of the sections of the destroyed 4th Unit. During this stage the melt's lamination occurred, followed by a break-through of the “lava” reservoir on the 11 th day of the accident and penetration of the “lava” into space under the reactor.
Secondary Uranium Minerals on the Surface of Chernobyl
“Lava”
Authors: Burakov B. E. , Strykanova E. E. , Anderson E. B.
Published: September 2012

The formation of uranium minerals is still continuing in Chernobyl Unit No.
4. Yellow products of alteration that stain the surface of Chernobyl “lava”
have been examined by SEM and X-ray diffraction methods. Secondary minerals
of uranium identified are: UO4·4H2O studtite;
UO3·2H2O epiianthinite;
UO2·CO3 rutherfordine; also,
Na4(UO2)(CO3)3 was identified
together with the sodium carbonate phases
Na3H(CO3)2·2H2O and
Na2CO3·H2O. These minerals formed due to
the interaction between fuel-containing masses or “lava”, water and air. The
matrices of the “lava” do not contain significant amounts of sodium. The
source of sodium may be water that has penetrated into the “Sarcophagus”.
All identified secondary minerals of uranium are highly unstable, and their
continued formation can seriously endanger the radiological situation of the
4th Unit.
Properties and Genesis of Hot Particles from the Chernobyl
Reactor Accident
Authors: Schubert-Bischoff Peter , Lutze Werner , Burakov Boris E.
Published: September 2012

On April 25, 1986, the nuclear reactor Unit 4 (RBMK) at Chernobyl, Ukraine,
exploded. Besides molecular species, the fallout contained particles of
relatively high specific activity (hot particles) with a wide range of
chemical compositions. The composition of a hot particle bears information
about its genesis. Particle sizes ranged from a few to 100s of micrometers.
Data on a hot particle, found in Berlin, Germany, is presented and discussed
in context with earlier measurements on other particles to understand their
genesis. The chemical composition was determined by electron probe micro
analysis. Our particles are either reactor fuel (one) or fission product
alloys (nine). The alloys were formed during normal reactor operation.
Strongly varying concentrations of Fe and Ni suggest that at least some of
our particles reacted with molten structural material of the reactor. The
particles were mobilized by fuel oxidation or fuel dust generation during
the accident. The fission product composition can only be explained if we
assume that the alloys remained in the solid state in the course of the
accident. Some particles may have been ejected during the explosion, others
later while the reactor was burning. Activities (103Ru and
106Ru, originally up to 160,000 Bq) of our ten year old
particles were re-measured but were no longer detectable. No long-lived
γ-emitters were found. The 99Tc activity was calculated and found
to only lBq. The γ -spectrum of the fuel particle still shows
137Cs (1 Bq) and 60Co (<1 Bq).
General Classification of “Hot” Particles from the Nearest
Chernobyl Contaminated Areas
Authors: Shabalev S. I. , Burakov B. E. , Anderson E. B.
Published: September 2012

The morphology and composition both chemical and radionuclide of the main
types of the solid-phase “hot” particles formed following the accident on
the Chernobyl NPP have been studied by SEM, electron microprobe and
gamma-spectrometry methods. Differences in many isotopes including:
106Ru, 134Cs, 137Cs dependent upon the
hot particle matrix chemical composition was observed. The classification of
hot particles based upon the chemical composition of their matrices has been
done. It includes three main types: 1) fuel particles with UOx
matrix; 2) fuel-constructional particles with Zr-U-0 matrix, 3) hot
particles with metallic inclusions of Fe-Cr-Ni. Moreover, there are more
rare types of hot particles with silicate or metal matrices. It was shown
that only metallic inclusions of Fe-Cr-Ni are concentrators of
106Ru, which caused this nuclides assimilation in the molten
stainless steel during the initial stages of the accident. Soils
contamination of non-radioactive lead oxide particles in the Chernobyl NPP
region were noticed. It was supposed that part of metallic lead, dropped
from helicopters into burning reactor during first days of accident, was
evaporated and oxidized accompanying solid oxide particles formation.
Dissolution Rates of DWPF Glasses from Long-Term
PCT
Authors: Ebert W. L. , Tam S.-W.
Published: September 2012

We have characterized the corrosion behavior of several Defense Waste
Processing Facility (DWPF) reference waste glasses by conducting static
dissolution tests with crushed glasses. Glass dissolution rates were
calculated from measured B concentrations in tests conducted for up to five
years. The dissolution rates of all glasses increased significantly after
certain alteration phases precipitated. Calculation of the dissolution rates
was complicated by the decrease in the available surface area as the glass
dissolves. We took the loss of surface area into account by modeling the
particles to be spheres, then extracting from the short-term test results
the dissolution rate corresponding to a linear decrease in the radius of
spherical particles. The measured extent of dissolution in tests conducted
for longer times was less than predicted with this linear dissolution model.
This indicates that advanced stages of corrosion are affected by another
process besides dissolution, which we believe to be associated with a
decrease in the precipitation rate of the alteration phases. These results
show that the dissolution rate measured soon after the formation of certain
alteration phases provides an upper limit for the long-term dissolution
rate, and can be used to determine a bounding value for the source term for
radionuclide release from waste glasses. The long-term dissolution rates
measured in tests at 20,000 m−1 at 90°C in tuff groundwater at pH
values near 12 are about 0.2,0.07, and 0.04 g/(m2•d) for the
Environmental Assessment glass and glasses made with SRL 131 and SRL 202
frits, respectively.
Simulation of Natural Corrosion by Vapor Hydration Test:
Seven-Year Results
Authors: Luo J. S. , Ebert W. L. , Mazer J. J. , Bates J. K.
Published: September 2012

We have investigated the alteration behavior of synthetic basalt and SRL 165
borosilicate waste glasses that had been reacted in water vapor at 70°C for
time periods up to seven years. The nature and extent of corrosion of
glasses have been determined by characterizing the reacted glass surface
with optical microscopy, scanning electron microscopy (SEM), transmission
electron microscopy (SEM), and energy dispersive x-ray spectroscopy (EDS).
Alteration in 70°C laboratory tests was compared to that which occurs at
150–200°C and also with Hawaiian basaltic glasses of 480 to 750 year old
subaerially altered in nature. Synthetic basalt and waste glasses, both
containing about 50 wt % SiO2, were found to react with water
vapor to form an amorphous hydrated gel that contained small amounts of
clay, nearly identical to palagonite layers formed on naturally altered
basaltic glass. This result implies that the corrosion reaction in nature
can be simulated with a vapor hydration test. These tests also provide a
means for measuring the corrosion kinetics, which are difficult to determine
by studying natural samples because alteration layers have often spalled off
the samples and we have only limited knowledge of the conditions under which
alteration occurred.
Isothermal Crystallization Kinetics in Simulated High-Level
Nuclear Waste Glass
Authors: Vienna John D. , Hrma Pavel , Smith Donald E.
Published: September 2012

Crystallization kinetics of a simulated high-level waste (HLW) glass were
measured and modelled. Kinetics of acmite growth in the standard HW39–4
glass were measured using the isothermal method. A
time-temperature-transformation (TTT) diagram was generated from these data.
Classical glass-crystal transformation kinetic models were empirically
applied to the crystallization data. These models adequately describe the
kinetics of crystallization in complex HLW glasses (i.e., RSquared = 0.908).
An approach to measurement, fitting, and use of TTT diagrams for prediction
of crystallinity in a HLW glass canister is proposed.
Further Development of the Jantzen - Plodinec Model of Glass
Durability
Authors: Shakhmatkin B. A. , Vedishcheva N. M. , Aloy A. S. , Plodinec M. J. , Marra S. L.
Published: September 2012

The hydration thermodynamic approach to the prediction of glass durability
was originally applied to nuclear waste glasses by Jantzen and Plodinec.
This approach is useful for control of the production of nuclear waste
glasses. However, improvements are necessary if the approach is to be
extended to different glasses, particularly those with higher alkali metal
concentrations. This is of special significance for vitrification of the
salt wastes at Hanford. Various methods for improving the predictive power
of the approach have been examined. Combining a more accurate representation
of the alkali metal species in the glass with a more rigorous thermodynamic
approach is a promising avenue to improved predictive power.
The Effect of Humic Acids on the Element Release from High Level Waste Glass
Authors: Wei J. , Van Iseghem P.
Published: September 2012

Eu and Am doped glasses were interacted with synthetic interstitial clay water (SIC) and corresponding reference leachant, humic acids free interstitial solution (IS) to investigate the influence of humic acids on the leaching behaviour of the waste glass. Static leach tests were carried out at 40°C and 90°C. The release of the lanthanide Eu and the actinide Am from the glass was obviously enhanced by the presence of humic acids. The leaching of transition elements, Fe and Ti strongly depends on the humic acids concentration. The leaching of glass matrix components, Al and B was also influenced by the concentrations of humic acids. However, humic acids have little effect on the leaching of glass matrix element Si.
Sintered Glasses for High-Level Wastes
Authors: Russo D. O. , Messi de Bernasconi N. B. , Sterba M. E. , Heredia A. D. , Sanfilippo M. , Prastalo S. , Sbriller J. C.
Published: September 2012

The objective of the work was to evaluate the long-term capacity of sintered
glass to retain high-level nuclear wastes (HLW) in near-repository
conditions. We have studied the corrosion behavior of waste forms partially
devitrified (43 vol.%) in different aqueous media. Devitrified samples were
irradiated at doses (γ radiation from a Co 60 source) ranging
from 1.4 × 106 Gy to 2.0 × 108 Gy, in order to study
their aqueous corrosion resistance in simulated underground water. The
results show little or no effect of irradiation on the density,
microstructure and corrosion resistance. The global dissolution rate was
almost constant around a value of 5×10−5 g. cm−2
d−1. Elemental dissolution rates were also unaffected by
radiation.
Effects of Redox Conditions of Water on Pu and Cm Leaching from Waste Glass
Authors: Inagaki Y. , Sakai A. , Furuya H. , Idemitsu K. , Arima T. , Banba T. , Maeda T. , Matsumoto S. , Tamura Y.
Published: September 2012

Static corrosion tests were performed on the simulated waste glass doped with Pu and Cm (PuO2; 0.22wt%, Cm2O3; 0.09%) in deionized water at 90 °C with S/V ratio of 2600 m−1under oxidizing and reducing conditions, respectively. The corrosion tests under oxidizing conditions were performed in air, while the corrosion tests under reducing conditions were performed in an airtight stainless steel container purged with mixed gas (Ar+5%H2), where the Eh of the solution was maintained at approximately -0.45 V vs SHE. The results showed that the redox conditions have no remarkable influence on the leaching behavior of Pu and Cm. It was also observed that the Pu and Cm concentrations in the 1.8nm filtered solutions were one or two orders of magnitude lower than those in the 450nm filtered solutions under both redox conditions, which suggests that the insoluble suspended fractions (colloidal particles) with the size from 1.8nm to 450nm are dominant phase for Pu and Cm in the solution under both redox conditions.
Waste Glass Leaching and Alteration Under Conditions of Open
Site Tests
Authors: Sobolev I. A. , Ojovan M. I. , Batyukhnova O.G , Ojovan N. V. , Scherbatova T. D.
Published: September 2012

The behaviour of waste glass was investigated under open site disposal
conditions. This glass was produced by vitrification of intermediate level
radioactive waste from nuclear power plants. Two types of borosilicate
glasses were obtained for two different reactor wastes, WWER and RBM.K.
Leaching and alteration mechanisms are discussed as well as the data
processing technique used for these long term tests. The decay of
radionuclides was accounted for in order to obtain correct results. The
leaching factors obtained can be used for the assesment of radionuclide
retention. Discontinuous leaching of Cs-137 has been observed during more
than 8 years testing time. The fluctuating leaching rate depends on glass
composition. The average leaching rate remains within (0.4 – 4) μg/sq.
sm·day.

Alteration of waste glass includes the formation of surface layers and
cracks on the glass surface. SEM analysis of glass was used to show these
surface layers. The thickness of the layers was determined to be within 2–6
μm. The structure of these layers depends on glass composition and the
interfacing environment.
A Direct, Single-Step Plasma ARC-Vitreous Ceramic Process for
Stabilizing Spent Nuclear Fuels, Sludges, and Associated
Wastes
Authors: Feng Xiangdong , Einziger Robert E. , Eschenbach Richard C.
Published: September 2012

A single-step plasma arc-vitreous ceramic (PAVC) process is described for
converting spent nuclear fuel (SNF), SNF sludges, and associated wastes into
a vitreous ceramic waste form. This proposed technology is built on
extensive experience of nuclear waste form development and nuclear waste
treatment using the commercially available plasma arc centrifugal (PAC)
system. SNF elements will be loaded directly into a PAC furnace with minimum
additives and converted into vitreous ceramics with up to 90 wt% waste
loading. The vitreous ceramic waste form should meet the functional
requirements for borosilicate glasses for permanent disposal in a geologic
repository and for interim storage. Criticality safety would be ensured
through the use of “batch” modes, and controlling the amount of fuel
processed in one batch. The minimum requirements on SNF characterization and
pretreatment, the one-step process, and minimum secondary waste generation
may reduce treatment duration, radiation exposure, and treatment cost.
Accelerated Testing of Waste forms using a Novel Pressurized
Unsaturated flow (PUF) Method
Authors: McGrail B. P. , Martin P. F. , Lindenmeier C. W.
Published: September 2012

A new experimental technique has been developed to test waste forms and
other proposed engineered-barrier materials under unsaturated conditions.
Laboratory experiments using the pressurized unsaturated flow (PUF)
apparatus have been performed with a Na-Ca-Al borosilicate glass being
studied for immobilization of low-activity tank wastes and a reference
borosilicate glass for immobilization of high-level wastes (SRL-202). A
complex coupling between glass corrosion, secondary phase precipitation, and
unsaturated flow behavior was observed. Precipitation of a family of
Na-Ca-Al zeolites was also found to cause an acceleration in the reaction
rate of the low-activity waste glass. This same effect was observed after
approximately 1 year in high solid-to-liquid ratio batch tests but after
only 12 days using the PUF method at the same temperature. The onset of
secondary phase precipitation was tracked by monitoring changes in
volumetric water content and by inline chemical analysis. Finally, elemental
release of the major glass components and solution pH were found to differ
under unsaturated flow-through conditions as compared with saturated, batch
tests. These findings, while preliminary, have important implications for
understanding and modeling glass corrosion behavior under unsaturated
conditions.
Nepheline Precipitation in High-Level Waste Glasses :
Compositional Effects and Impact on the Waste Form
Acceptability
Authors: Li H. , Vienna J. D. , Hrma P. , Smith D. E. , Schweiger M. J.
Published: September 2012

The impact of crystalline phase precipitation in glass during canister
cooling on chemical durability of the waste form limits waste loading in
glass, especially for vitrification of certain high-level waste (HLW)
streams rich in Na2O and Al2O3. This study
investigates compositional effects on nepheline precipitation in simulated
Hanford HLW glasses during canister centerline cooling (CCC) heat treatment.
It has been demonstrated that the nepheline primary phase field defined by
the Na2O-Al2O3-SiO2 ternary
system can be used as an indicator for screening HLW glass compositions that
are prone to nepheline formation. Based on the CCC results, the component
effects on increasing nepheline precipitation can be approximately ranked as
Al2O3 > Na2O > Li2O ≈
K2O ≈ Fe2O3 > CaO >
SiC2. The presence of nepheline in glass is usually
detrimental to chemical durability. Using x-ray diffraction data in
conjunction with a mass balance and a second-order mixture model for 7-day
product consistency test (PCT) normalized B release, the effect of glass
crystallization on glass durability can be predicted with an uncertainty
less than 50% if the residual glass composition is within the range of the
PCT model.
Phase Separation in Simulated Plutonium Glasses with Phosphate and Fluorine and the effect on Glass Corrosion in Water
Authors: Li H. , Vienna J. D. , Chen Y. L. , Wang L. Q. , Liu J.
Published: September 2012

The solubility limit of phosphate in glass was found to decrease as fluorine increased. Amorphous phase separation was found in the glass as a result of interaction between phosphate and fluorine, in which Ce and Gd strongly partitioned. The glasses exhibiting amorphous phase separation showed higher normalized B releases according to the 31-day product consistency test (PCT) compared to glasses without phase separation. Phosphate that leached from glass into water increased the concentrations of Ce and Gd in the 7- and 31-day PCT solutions by more than an order of magnitude. The liquid state31P-NMR results suggested that phosphate in water interacts with Ce or Gd. Therefore, the observed concentration jump for Ce and Gd in the PCT solution may be attributed to the increases in the solubility limits of Ce and Gd as a result of phosphate complexation with Ce and Gd.
Chemistry and Performance of Blended Cements and Backfills for use in Radioactive Waste Disposal
Authors: Duerden S. , Glasser F. P. , Goldthorpe K. , Pedersen J. , Quillin K. , Ross D. , Stronach S. A. , Tyrer M.
Published: September 2012

The ability of NaCl and MgSO4to impair the performance of Portland cement, blended cements containing slag and fly ash and of a permeable backfill have been measured. Performance is determined by decrease in pH, changes in mineralogy and loss of physical coherence. Experiments have been made at 25°, 55° and 85°C and extensively backed up by chemical models of cement performance. NaCl, up to 1.5M, has a comparatively slight impact on performance but MgSO4rapidly and almost quantitatively reacts, lowering system pH's to < 10, conditioned by mixtures of Mg(OH)2and magnesium silicates with gypsum.
Characterization of Solidified Radioactive Waste Due to the
Incorporation of High-and Low-Density Polyethylene Granules and Titanium
Dioxide in Mortar Matrices
Authors: Perić Aleksandar
Published: September 2012

The rutile form of titanium dioxide (TiO2) and granules of high
density polyethylene (PEHD) and low density polyethylene (PELD) were used to
prepare mortar matrices for immobilization of radioactive waste materials
containing ‘137Cs. PELD,PEHD and TiO2 were added to
mortar matrix preparations with the objective of improving physico-chemical
characteristics of the radwaste-mortar matrix mixtures, in particular the
leach-rate of the immobilized radionuclide. The diameters of the PELD and
PEHD used varied from 0.2 to 2.0 mm. One type of PELD and two types of PEHD
were used to replace 50 weight percent of stone granules, average diameter
of 2 mm, normally used in the matrix, in order to decrease the porosity and
density of the mortar matrix and to avoid segregation of the stone particles
at the bottom of the immobilized radioactive waste cylindrical form.
TiO2 was also added to the mortar formulation, replacing 5 and
8 weight percent of the total cement weight, for each PEHD and PELD
formulation. Cured samples were investigated under temperature stress
conditions, where the temperature extremes were: Tmill =
-20°C,Tmax= +70°C. Samples were periodically immersed in
distilled water at the ambient room temperature, after each freezing and
heating treatment. Results of accelerated leaching experiments for these
samples and samples prepared exclusively with polyethylenes replacing 100
percents of the stone granules and TiO2, treated in
nonaccelerated leaching experiments, were compared. Even using an
accelerated ageing leach test that overestimates 137Cs leach
rates, it can be deduced, that radionuclide leach rates from the radioactive
waste mortar mixture forms were improved. Leach rates decreased from 5
percent, for the material prepared with stone aggregate, to 3.1 to 4.0
percent, for the materials prepared solely with PEHD, PELD or
TiO2, and to about 3 percents for all six types of the
TiO2-PEHD and TiO2-PELD mixtures tested.
Dissolution of Synroc in Deionised Water at
150°C
Authors: Smith Katherine L. , Colella Michael , Thorogood Gordon J. , Blackford Mark G. , Lumpkin Gregory R. , Hart Kaye P. , Prince Kathryn , Loi Elaine , Jostsons Adam
Published: September 2012

Synroc containing 20 wt% simulated high level waste (HLW) was subjected to
two sets of leach tests at 150°C where the leachant was and was not replaced
during the test (replacement and non-replacement testing). The leachant was
a KH-phthalate buffered solution (pH 4.2). Samples were characterised before
and after leach testing using SEM, AEM and SIMS. Elemental concentrations in
leachates were measured using ICP-MS. In concert with the findings of i) a
dissolution study of perovskite in a flowing leachant and ii) a previous
Synroc dissolution study (wherein Synroc containing 10 wt% simulated HLW was
subjected to periodic replacement, leach testing in deionised water at
150°C), the results of this study show that when the leachant replacement
frequency is varied from 7 d to the duration of the test, there is no effect
on leach rate or leaching mechanisms.
Alteration of Cold Crucible Melter Titanate-Based Ceramics:
Comparison with Hot-Pressed Titanate-Based Ceramic
Authors: Advocat T. , Leturcq G. , Lacombe J. , Berger G. , Day R. A. , Hart K. , Vernaz E. , Bonnetier A.
Published: September 2012

Synroc ceramics were synthesized in an induction-heated cold crucible at
laboratory scale (1 kg) from an oxide mixture, and at industrial prototype
scale (45 kg) from Synroc previously produced by sintering under load at
high temperature. After melting, both materials contained the major phases
of Synroc-C. The chemical durability of both melted materials, as determined
by static leaching of powder samples in initially pure water at 90°C with an
SA/V ratio of 20000m−1, was equivalent to that of conventional
hot-pressed Synroc-C. Cerium, used in this investigation to simulate the
presence of tri-and tetravalent actinides, was found in steady-state
concentrations on the order of 1 ppb (i.e. NL(Ce) ≤ 10−6
g·m−2). The concentration in the leachates was independent of
the initial CeO2 content of the Synroc (at least up to 10 wt%);
moreover, it is similar to the results obtained with hot-pressed Synroc-C
specifically formulated for conditioning long-lived actinides.
Study of Melted Synroc Doped with Simulated High-Level
Waste
Authors: Sobolev I. A. , Stefanovsky S. V. , Ioudintsev S. V. , Nikonov B. S. , Omelianenko B. I. , Mokhov A. V.
Published: September 2012

Preparation and characterization of inductively-melted Synroc containing 20
wt% simulated plant “Mayak” reprocessing waste were performed. The sample
bulk composition was as follows, (in wt.%): 55.4 TiO2; 15.8
ZrO2; 7.5 CaO; 7.4 BaO; 4.3 Al2O3 2.0
MnO; 1.8 SiO2; 0.7 Na2O; 1.9 K2O, 0.5
Ce2O3; 1.0 UO2; 0.9 NiO; 0.6
Cr2O3, and 0.2 FeO. The sample was produced by
melting in air at 1550–1600 °C under barometric pressure. It is composed of
a few crystalline phases and a minor glass phase. Most of the phases
(hollandite, zirconolite, perovskite and rutile) are similar to the
analogous phases found in the other Synroc formulations. An additional phase
with average composition, wt.%: 59.8 TiO2; 15.6 CaO; 7.0
UO2; 5.6 ZrO2; 4.7 MnO; 4.1
Ce2O3, and 1.8 Al2O3 was
found. Some elements (Ba, Si, Ni, K, Na, Fe) were present in the phase in
negligible quantities. Its formula
(Ca2.65U0.3Ce0.2)(Ti7.3Mn0.6Zr0.4Al0.3)O20.0
is rather close to a rare mineral uhligite -
Ca3(Ti,Zr,Al)9O20. Another possible
counterpart of the phase is murataite-like mineral previously described in
tailored ceramic designed for Savannah River Plant wastes fixation. This
phase as well as zirconolite are the major host for U in the sample
Preliminary data on the material leachability in water at 350 °C and 50 MPa
have been obtained Uranium contents in the solution were about 1 ppb and
close to the uranium dioxide solubility in deionized water under the same
P-T conditions.
Comparative Study of Synroc-C Ceramics Produced by Hot-Pressing
and Inductive Melting
Authors: Sobolev I. A. , Stefanovsky S. V. , Omelianenko B. I. , Ioudintsev S. V. , Vance E. R. , Jostsons A.
Published: September 2012

Three Synroc-C samples, containing simulated high level waste were studied.
One was produced by the conventional hot-pressing method at ANSTO,
Australia, and the others were obtained using cold crucible technology at
Radon, Russia. One of the melted samples was prepared using the Australian
sol-gel precursor and the second one was obtained from an oxide-nitrate
mixture. It was established that the specimens have closely similar mineral
compositions, with major hollandite, perovskite, zirconolite, and rutile.
Small amounts of hibonite were also found. Unlike the hot-pressed Synroc
containing metallic alloy particles, melted Synrocs contain molybdates. An
investigation of mineral compositions and elemental distribution in the
samples was carried out. Features of hot-pressed and melted ceramics were
compared. Unit cell parameters of the Synroc phases were determined and
preliminary results on durability of the melted Synroc are presented.