bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

ABSTRACTThe 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

ABSTRACTSynroc 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

ABSTRACTSynroc 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

ABSTRACTPreparation 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

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

ABSTRACTSyntheses 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

ABSTRACTThe 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

ABSTRACTZirconolite-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

ABSTRACTAn 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

ABSTRACTSeveral 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

ABSTRACTThe 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

ABSTRACTChlorine-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

ABSTRACTA 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

ABSTRACTThe 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

ABSTRACTExtensive 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

ABSTRACTCCIM 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

ABSTRACTThe 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

ABSTRACTThe 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.