pH-Dependency of HLW Glass Dissolution and Sorption Behavior of Eu, Th, and U on Glass Corrosion Products

1997 ◽  
Vol 506 ◽  
Author(s):  
B. Luckscheiter ◽  
B. Grambow
Keyword(s):  
Dissolution Rate ◽  
Sorption Behavior ◽  
Dissolved Silica ◽  
First Order ◽  
Glass Dissolution ◽  
Glass Corrosion ◽  
Ph Dependency ◽  
Constant Rate ◽  
Radionuclide Release ◽  
Experimental Release

ABSTRACTRadionuclide release rates of HLW glasses in brines are normally orders of magnitude lower than glass corrosion rates. Various mechanisms were invoked to explain the experimental release; they may be controlled by the glass dissolution rate, solubility, sorption, coprecipitation, etc.. Glass matrix corrosion kinetics is described by a first order equation, where pH, S/V ratio, an initial constant rate and affinity term based on orthosilic acid activity are the key parameters. The dissolution rate decreases significantly as dissolved silica accumulates in solution

scholarly journals Dissolution Rates of DWPF Glasses from Long-Term PCT

1996 ◽  
Vol 465 ◽  
Author(s):  
W. L. Ebert ◽  
S.-W. Tam
Keyword(s):  
Dissolution Rate ◽  
Surface Area ◽  
Spherical Particles ◽  
Dissolution Rates ◽  
Glass Dissolution ◽  
Dissolution Tests ◽  
Dissolution Model ◽  
Advanced Stages ◽  
Radionuclide Release

ABSTRACTWe 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.

Modeling The Dissolution Behavior of Defense Waste Glass in A Salt Repository Environment

1987 ◽  
Vol 112 ◽  
Author(s):  
B. P. McGrail
Keyword(s):  
Mechanistic Model ◽  
Waste Glass ◽  
Dissolution Behavior ◽  
Precipitation Reaction ◽  
Iron Corrosion ◽  
Silicate Mineral ◽  
Surface Area Ratio ◽  
Dissolved Silica ◽  
Glass Dissolution ◽  
Radionuclide Release

AbstractA mechanistic model describing a dynamic mass balance between the production and consumption of dissolved silica was found to describe the dissolution behavior of SRL-165 defense waste glass in a high-magnesium brine (PBB3) at a temperature of 90°C. The synergistic effect of the waste package container on the glass dissolution rate was found to depend on a precipitation reaction for a ferrous silicate mineral.The model predicted that the ferrous silicate precipitate should be variable in composition where the iron/silica stoichiometry depended on the metal/glass surface area ratio used in the experiment. This prediction was confirmed experimentally by the variable iron/silica ratios observed in filtered leachates. However, the interaction between dissolved silica and iron corrosion products needs to be much better understood before the model can be used with confidence in predicting radionuclide release rates for a salt repository.

scholarly journals The fate of Si and Fe while nuclear glass alters with steel and clay

2021 ◽  
Vol 5 (1) ◽  
Author(s):  
C. Carriere ◽  
P. Dillmann ◽  
S. Gin ◽  
D. Neff ◽  
L. Gentaz ◽  
...  
Keyword(s):  
Dissolution Rate ◽  
Protective Layer ◽  
Glass Dissolution ◽  
A Value ◽  
Waste Forms ◽  
Controlling Mechanism ◽  
Geological Repository ◽  
The Mean ◽  
High Level ◽  
Initial Dissolution Rate

AbstractThe French concept developed to dispose high-level radioactive waste in geological repository relies on glassy waste forms, isolated from the claystone host rock by steel containers. Understanding interactions between glass and surrounding materials is key for assessing the performance of a such system. Here, isotopically tagged SON68 glass, steel and claystone were studied through an integrated mockup conducted at 50 °C for 2.5 years. Post-mortem analyses were performed from nanometric to millimetric scales using TEM, STXM, ToF-SIMS and SEM techniques. The glass alteration layer consisted of a crystallized Fe-rich smectite mineral, close to nontronite, supporting a dissolution/reprecipitation controlling mechanism for glass alteration. The mean glass dissolution rate ranged between 1.6 × 10−2 g m−2 d−1 to 3.0 × 10−2 g m−2 d−1, a value only 3–5 times lower than the initial dissolution rate. Thermodynamic calculations highlighted a competition between nontronite and protective gel, explaining why in the present conditions the formation of a protective layer is prevented.

Measurement of Glass Corrosion in Boom Clay Disposal Conditions: First Results of the Experimental Programme 2000-2003 of SCK•CEN

2003 ◽  
Author(s):  
Karel Lemmens ◽  
Marc Aertsens ◽  
Ve´ra Pirlet ◽  
Norbert Maes ◽  
Hugo Moors ◽  
...  
Keyword(s):  
Low Temperature ◽  
Dissolution Rate ◽  
Near Field ◽  
Glass Frit ◽  
Glass Mass ◽  
Boom Clay ◽  
Glass Dissolution ◽  
Mass Losses ◽  
Dissolution Behaviour ◽  
First Results

To estimate the lifetime of vitrified high level waste (HLW-glass) in Boom Clay disposal conditions, the dissolution behaviour of waste glass has been studied with experiments performed in surface laboratories and in the HADES underground research facility of SCK·CEN since 1980. We present the main topics and first results of the SCK·CEN programme 2000–2003. This programme focuses on the following items: (1) the diffusion/sorption/precipitation of silica in Boom clay or backfill clay, (2) demonstration of glass dissolution behaviour in realistic test conditions, (3) the effect of presaturation of the clay with silica, and (4) the estimation of near field concentrations of critical isotopes. The experiments have shown so far that Si, released by the glass, is effectively immobilized by Boom Clay, but it can nevertheless diffuse into the clay without immediately precipitating. The dissolution rate of glass SON68 and SM539 is determined in Boom Clay at in situ density and at 30°C (this is the long-term temperature expected near the waste glass packages in a Boom Clay repository). The dissolution rates, based on glass mass losses, are constant during the first year, at ∼ 0.010 g.m−2.day−1 for glass SON68 and ∼ 0.012 g.m−2.day−1 for glass SM539. The addition of glass frit causes a decrease of the glass dissolution rate, both with glass SON68 and SM539, and both in Boom Clay and in FoCa-clay. In FoCa-clay at high density with glass frit, the dissolution rates, based on glass mass losses, after 8 months at 30°C are ∼ 0.001 g.m−2.day−1 (SM539) and ∼0.005 g.m−2.day−1 (SON68). Because the experiments performed in Boom Clay and FoCa-clay with glass frit simulate realistic conditions (high clay density, low temperature), they can be used to estimate the maximum glass dissolution rate in a (Boom) clay repository. The corresponding minimum lifetime of a glass canister, calculated with the SCK·CEN code for lifetime predictions, is of the order of 105 to 106 years, if we neglect the internal glass surface area (due to cracking). In more diluted clay suspensions with glass frit, the glass dissolution rate is 10−4 to 10−5 g.m−2.day−1 or even zero. This would correspond to a lifetime of >>106 years. So far, there is no indication that the addition of glass frit leads to secondary phase formation at low temperature (30–40°C). Leach experiments with doped glasses SON68 and SM539 suggest that the maximum concentrations of most critical radionuclides in near field conditions are lower than the best estimate solubilities used for performance assessment studies in Boom Clay. For Se, relatively high concentrations were measured, though. The research programme for the underground laboratory is not discussed.

Diffusion of Radionuclides in Compacted Bentonite : Results from Combined Glass Dissolution and Migration Tests

1994 ◽  
Vol 353 ◽  
Author(s):  
Masaki Tsukamoto ◽  
T. Ohe ◽  
T. Fujita ◽  
R. HesbÖl ◽  
H-P. Hermansson
Keyword(s):  
Dry Density ◽  
Apparent Diffusion Coefficients ◽  
Glass Dissolution ◽  
Glass Corrosion ◽  
Gas Atmosphere ◽  
Apparent Diffusion ◽  
Calculation Results ◽  
The Difference ◽  
And Migration ◽  
Elemental Mass

AbstractDiffusion experiments of radionuclides in compacted sodium bentonite with a dry density of 1.0 g/cm3 were performed in nitrogen gas atmosphere at 90 °C for 208 d and 375 d. The corrosion experiments of crushed radioactive glass, JSS-A, carried out simultaneously to provide the source of the radionuclides for the diffusion experiments. The normalized elemental mass losses of cesium isotopes and 238Pu were lower than those of boron (ca. 10 g/m2) probably because of the difference of sorption and/or precipitation. The apparent diffusion coefficients of 238Pu, 234U and 125Sb were determined to be 2x 10-14 m2/s, 5x 10-12 m2/s and 2x 10-12 m2/s, respectively. The distribution coefficient of Pu estimated from the diffusion data was of the same order as that from batch sorption experiments. The glass corrosion and the plutonium diffusion were described by the geochemical codes PHREEQE, STRAG4 and GESPER. The calculation results well fitted the observed data.

PENYERAPAN GAS KARBONDIOKSIDA (CO 2 ) DALAM BIOGAS DENGAN LARUTAN Ca(OH)2

EKUILIBIUM ◽  
2015 ◽  
Vol 14 (1) ◽  
Author(s):  
Paryanto Paryanto
Keyword(s):  
Experimental Data ◽  
Carbon Dioxide ◽  
Order Reaction ◽  
First Order Reaction ◽  
Glass Ball ◽  
Carbon Dioxide Content ◽  
First Order ◽  
Constant Rate ◽  
Rate Of Reaction ◽  
Reduce Carbon Dioxide

<p>Abstract: Carbon dioxide content in biogas produced by fermentation is still high. Because of<br />that, biogas need a process purification to reduce carbon dioxide content. In this work, Ca(OH)2<br />solution was contacted with biogas in a column for reducing the CO<br />2<br />content. This research<br />studied the effect of packing type used in absorber column on the rate of CO<br />2<br />reduction. Based<br />on experimental data and modelling, it was found that the reaction between CO<br />2<br />followed first order reaction. The constant of rate reaction was affected by the packing type<br />which using glass ball, plastic pipe, ceramic, wood, and clay roof, the constant rate were 0.781,<br />0.464, 0.916, 0.637, and 0.987 min<br />Keywords: Biogas, CO<br />2<br />, Ca(OH)2<br />-1<br />, respectively.<br />, absorber, rate of reaction</p>

Dissolution Kinetics of a Simple Analogue Nuclear Waste Glass as a Function of Ph, Time and Temperature

1989 ◽  
Vol 176 ◽  
Author(s):  
Kevin G. Knauss ◽  
William L. Bourcier ◽  
Kevin D. McKeegan ◽  
Celia I. Merzbacher ◽  
Son N. Nguyen ◽  
...  
Keyword(s):  
Dissolution Rate ◽  
Rate Constant ◽  
Rate Constants ◽  
Dissolution Kinetics ◽  
State Theory ◽  
Alkaline Solutions ◽  
Glass Dissolution ◽  
Constant Ph ◽  
Transport Control ◽  
Dissolution Rate Constants

ABSTRACTWe have measured the dissolution rate of a simple five-component borosilicate glass (Na2O, CaO, Al2O3, B2O3, SiO2) using a flow-through system. The experiments were designed to measure the dissolution rate constant over the interval pH 1 through pH 13 at 3 temperatures (25°, 50° and 70°C). Dilute buffers were used to maintain a constant pH. Analyses of solutions and solid surfaces provided information that is used to develop a kinetic model for glass dissolution.Under all conditions we eventually observed linear dissolution kinetics. In strongly acidic solutions (pH 1 to pH 3) all components but Si were released in their stoichiometric proportions and a thick, Si-rich gel was formed. In mildly acidic to neutral solutions the gel was thinner and was both Si- and Al-rich, while the other components were released to solution in stoichiometric proportions. In mildly to strongly alkaline solutions all components were released to solution in stoichiometric proportions. By varying the flow rate at each pH we demonstrated a lack of transport control of the dissolution rate.The dissolution rates were found to be lowest at near-neutral pH and to increase at both low and high pH. A rate equation based on transition-state theory (TST) was used to calculate dissolution rate constants and reaction order with respect to pH over two pH intervals at each temperature. At 250C between pH 1 and pH 7 based on the Si release rate the log rate constant for glass dissolution (g glass/m20d) was −0.77 and the order with respect to pH was −0.48. Between pH 7 and pH 13 the log rate constant for glass dissolution was −8.1 and the order with respect to pH was +0.51. The measured simple glass dissolution rate constants compare very well with constants estimated by fitting the same TST equation to experimental results obtained for SRL-165 glass and to dissolution rate estimates made for synthetic basaltic glasses.

Oxidation of GR-S Vulcanizates

1946 ◽  
Vol 19 (3) ◽  
pp. 696-711
Author(s):  
J. Reid Shelton ◽  
Hugh Winn
Keyword(s):  
General Nature ◽  
Oxygen Absorption ◽  
Specimen Thickness ◽  
Absorption Data ◽  
First Order ◽  
Rapid Absorption ◽  
Constant Rate ◽  
Absorption Studies ◽  
Three Stages ◽  
Useful Life

Abstract Oxygen absorption studies on a tread type GR-S vulcanizate show: 1. There are three stages involved in the oxidation: (1) an initial rapid absorption, of apparent first order at constant pressure, which is of short duration and appears to involve a limited number of especially reactive centers in the vulcanizate ; (2) a constant-rate reaction, of apparent zero order at constant oxygen pressure, which would extend over most of the useful life of the vulcanizate in service ; (3) an autocatalytic reaction, beginning after a given stock has absorbed a definite amount of oxygen that is independent of the temperature at which the absorption takes place. 2. The rate of oxygen absorption increases with temperature. The temperature coefficient for the constant rate reaction is 2.6 for a 10° C change in temperature over the range 80° to 120° C. 3. Diffusion is not rate controlling, provided the specimen thickness does not exceed 0.080 inch at 100° C or 0.040 inch at 120° C. 4. The effect of cure on the rate of oxygen absorption by GR-S vulcanizates is slight, and consequently, the technique is particularly useful for studying the effect of changes in compounding upon aging resistance. 5. Oxygen absorption data can be satisfactorily correlated with changes in physical properties during aging. While the fundamental nature of the mechanism of the oxidation of GR-S is still unknown, considerable information has been obtained with respect to some of the factors involved, and the general nature of the reaction has been established. This information, together with the work still in progress, should be of value in pointing the way toward possible methods of improving the resistance of GR-S to aging.

Remaining Uncertainties in Predicting Long-Term Performance of Nuclear Waste Glass From Experiments

1993 ◽  
Vol 333 ◽  
Author(s):  
B. Grambow ◽  
Kernforschungszentrum Karlsruhe
Keyword(s):  
Current Knowledge ◽  
Research Work ◽  
Waste Glass ◽  
Dissolution Mechanism ◽  
Glass Dissolution ◽  
Radionuclide Release ◽  
Long Term Performance ◽  
Term Performance ◽  
Term Maintenance

ABSTRACTThe current knowledge on the glass dissolution mechanism and the representation of glass dissolution concepts within overall repository performance assessment models are briefly summarized and uncertainties related to mechanism, radionuclide chemistry and parameters are discussed. Understanding of the major glass dissolution processes has been significantly increased in recent years. Long-term glass stability is related to the long-term maintenance of silica saturated conditions. The behavior of individual radionuclides in the presence of a dissolving glass has not been sufficiently and results do not yet allow meaningful predictions. Conservative long-term predictions of glass matrix dissolution as upper limit for radionuclide release can be made with sufficient confidence, however these estimations generally result in a situation were the barrier function of the glass is masked by the efficiency of the geologic barrier. Realistic long-term predictions may show that the borosilicate waste glass contributes to overall repository safety to a much larger extent than indicated by overconservatism. Today realistic predictions remain highly uncertain and much more research work is necessary. In particular the long-term rate under silica saturated conditions needs to be understood and the behavior of individual radionuclides in the presence of a dissolving glass deserves more systematic investigations.

Reactions in the System Basalt/Simulated Spent Fuel/Water

1983 ◽  
Vol 26 ◽  
Author(s):  
D.E. Grandstaff ◽  
G.L. Mckeon ◽  
E.L. Moore ◽  
G.C. Ulmer
Keyword(s):  
Water System ◽  
Potassium Feldspar ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Glass Dissolution ◽  
Ph Values ◽  
C 30 ◽  
High Level ◽  
Radionuclide Release

ABSTRACTThe Grande Ronde Basalts underlying the Hanford Site are being evaluated as a possible site for a high-level nuclear waste repository. Experiments, in which basalt from the Umtanun flow of the Grande Ronde Basalt and basalt with simulated spent fuel were reacted with synthetic Hanford groundwater, were conducted to determine steady state concentrations which can be used in radionuclide release-rate models. Tests were performed at temperatures of 100°, 200°, and 300°C; 30 MPa pressure, and a solution:solid mass ratio of 10:1 for durations up to 7,000 hr. Solution aliquots were extracted periodically during the experiments for analysis. The pH was measured at 250°C and recalculated to higher temperatures. In the basalt-water system the stable high-temperature pH values achieved were 7.2 (100°C), 7.5 (200°C), and 7.6 (300°C). Solution composition variations are due to mesostasis (glass) dissolution and precipitation of secondary phases. Solution measurements indicate a redox potential (Eh) of about -0.7 volts at 300°C. Secondary phases produced include silica, potassium feldspar, iron oxides, clays, scapolite, and zeolites. Tests in the basalt + simulated spent fuel + water systen show that calculated pH values stabilized near 7.6 (100°C), 7.2 (200°C), and 7.7 (300°C). At higher temperatures, solution concentrations were controlled by secondary phases similar to those found in basalt-water tests. Less than 1% of uranium, thorium, samarium, rhenium, cerium, and palladium were released to solution while somewhat higher amounts of iodine, molybdenum, and cesium were released. The UO2 component was unreactive; however, other components (e.g., cesium-bearing phases) were almost completely dissolved. Secondary phases incorporating radionuclide-analog elements include clays, palladium sulfide, powellite, coffinite, and a potassium-uranium silicate.

Sign in / Sign up

Export Citation Format

Share Document