Simulation of the Waste Glass Behavior in a Loamy Soil of the Wet Repository Site

2000 ◽  
Vol 663 ◽  
Author(s):  
M.I. Ojovan ◽  
N.V. Ojovan ◽  
I.V. Startceva ◽  
G.N. Chuikova ◽  
A.S. Barinov
Keyword(s):  
Borosilicate Glass ◽  
Field Data ◽  
Waste Glass ◽  
Loamy Soil ◽  
Model Calculations ◽  
Near Surface ◽  
Glass Corrosion ◽  
Radionuclide Release ◽  
Glass Behavior

ABSTRACTA model developed for description of waste glass corrosion has been applied to assess the radionuclide release from real radioactive (intermediate level) vitrified material over extended storage periods. Field data generated during the long-term testing of the prototype waste glass packages were mathematically processed and the derived parameters used in model calculations. Regardless of the corrosive saturated conditions of the wet near-surface repository, the fairly high safety of trench disposal has been demonstrated for borosilicate glass containing real NPP- operational waste.

Modelling of the Waste Form Behaviour in a Wet Near-Surface Repository Site Over Extended Time Periods

2001 ◽  
Author(s):  
Michael I. Ojovan ◽  
Natalia V. Ojóvan ◽  
Irene V. Startceva ◽  
Zoja I. Golubeva ◽  
Alexander S. Barinov
Keyword(s):  
Experimental Data ◽  
Surface Area ◽  
Salt Content ◽  
Volume Ratio ◽  
Waste Glass ◽  
Near Surface ◽  
Waste Forms ◽  
Time Periods ◽  
Radionuclide Release

Abstract A mathematical model was used to predict radionuclide release from bitumen and glass waste forms over extended time periods. To calculate some model parameters, we used experimental data derived from 12yr field tests with six borosilicate waste glass blocks (each ∼30 kg in weight) and a bitumen block (310 kg), containing real intermediate-level NPP operational waste (NaNO3, 86 wt.% of a dry salt content; 137Cs, 82% of the radioactive inventory). Specific radioactivities of the glass material containing 35 wt.% waste oxides were βtot(90Sr+90Y), 3.74×106 Bq/kg, and αtot(239Pu), 1.3×104Bq/kg. The bitumen block with ∼31 wt.% salt content and βtot(90Sr+90Y), 4.0·106 Bq/kg, and αtot(239Pu), 3.0×103 Bq/kg was manufactured on base of a hard bitumen BN-IV. Tests with the waste forms were performed under saturated conditions of an experimental near-surface repository with a free access of groundwater to the waste blocks through a covering of host loamy soil and backfill of coarse sand. The way used to quantify the amount of leached radioactivity was to measure the volume and radioactivity concentrations of contacting groundwater. In the model, radionuclide release from the waste glass is assumed to be controlled by the processes of diffusion limited ion exchange and glass network dissolution. The mechanism of radionuclide release from the bitumen matrix is believed to remain the same throughout the long-term storage period, except for the initial stage when an enhanced leaching from the surface layer occurs. This long-term release is assumed to be controlled by diffusion of radionuclides through the bitumen matrix. So, identical formulae were applied to calculate the values of leached radioactivity fractions for two waste forms. Radioactivity release curves were plotted for field data and calculation results. For both waste forms, there was good agreement between the modelled and available experimental data. According to the modelling results, fmax = 2.3×10−3% of the initial radioactivity will release from the waste glass into the environment within a proposed institutional control period of 300 years under conditions of the near-surface repository and in the absence of additional engineered barriers. For the bitumen block and the same 300-yr period, the total (maximum) leached radioactivity fraction will be fmax = 4.2×10−3%. The main result of the modelling and experimental studies concerning the leaching behaviour of the bituminised and vitrified waste materials is that the fractional radioactivity release for two waste forms is on the same order of magnitude. Numerical release values per a unit of a surface area to volume ratio are also rather close for two waste forms (exposed surface area to volume ratio for the bitumen block is 2 to 4 times greater then for the glass).

Leaching Chemistry of Defense Borosilicate Glass

1982 ◽  
Vol 15 ◽  
Author(s):  
Richard M. Wallace ◽  
George G. Wicks
Keyword(s):  
Predictive Model ◽  
Borosilicate Glass ◽  
Waste Glass ◽  
Savannah River ◽  
Short Term ◽  
Glass Corrosion ◽  
Savannah River Plant ◽  
Plant Waste ◽  
Long Term Behavior

Studies of the leachability of waste glass have been in progress at Savannah River Laboratory (SRL) for several years. The principal objective of these studies has been to predict the long-term behavior of Savannah River Plant waste glass when stored in a repository. Such predictions can be made from the results of short-term tests only if the mechanisms of waste glass corrosion are understood. Determining the mechanisms of corrosion and developing a predictive model have therefore been a major thrust of our work.

Waste glass behavior in a loamy soil of a wet repository site

2001 ◽  
Vol 298 (1-2) ◽  
pp. 174-179 ◽  
Author(s):  
M.I Ojovan ◽  
N.V Ojovan ◽  
I.V Startceva ◽  
G.N Tchuikova ◽  
Z.I Golubeva ◽  
...  
Keyword(s):  
Waste Glass ◽  
Loamy Soil ◽  
Glass Behavior

scholarly journals Laboratory Testing of Waste Glass Aqueous Corrosion; Effects of Experimental Parameters

1993 ◽  
Vol 333 ◽  
Author(s):  
W.L. Ebert ◽  
J.J. Mazer
Keyword(s):  
Ion Exchange ◽  
Corrosion Rate ◽  
Laboratory Tests ◽  
Water Diffusion ◽  
Waste Glass ◽  
Solution Chemistry ◽  
Exchange Reactions ◽  
Glass Corrosion ◽  
Hydrolysis Reactions

ABSTRACTA literature survey has been performed to assess the effects of the temperature, glass surface area/leachate volume ratio, leachant composition, leachant flow rate, and glass composition (actual radioactive vs. simulated glass) used in laboratory tests on the measured glass reaction rate. The effects of these parameters must be accounted for in mechanistic models used to project glass durability over long times. Test parameters can also be used to highlight particular processes in laboratory tests. Waste glass corrosion results as water diffusion, ion exchange, and hydrolysis reactions occur simultaneously to devitrify the glass and release soluble glass components into solution. The rates of these processes are interrelated by the effects of the solution chemistry and glass alteration phases on each process, and the dominant (fastest) process may change as the reaction progresses. Transport of components from the release sites into solution may also affect the observed corrosion rate. The reaction temperature will affect the rate of each process, while other parameters will affect the solution chemistry and the particular processes that are observed during the test. The early stages of corrosion will be observed under test conditions which maintain dilute leachates and the later stages will be observed under conditions that generate more concentrated leachate solutions. Typically, water diffusion and ion exchange reactions dominate the observed glass corrosion in dilute solutions, while hydrolysis reactions are dominant in more concentrated solutions. Which process controls the long-term glass corrosion is not fully understood, and the long-term corrosion rate may be either transport- or reaction-limited.

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.

scholarly journals Applying laboratory methods for durability assessment of vitrified material to archaeological samples

2021 ◽  
Vol 5 (1) ◽  
Author(s):  
Lorena Nava-Farias ◽  
James J. Neeway ◽  
Michael J. Schweiger ◽  
José Marcial ◽  
Nathan L. Canfield ◽  
...  
Keyword(s):  
Nuclear Waste ◽  
Iron Age ◽  
Test Methods ◽  
Durability Test ◽  
Laboratory Methods ◽  
Near Surface ◽  
Glass Corrosion ◽  
Durability Assessment ◽  
Nuclear Waste Forms

AbstractLaboratory testing used to assess the long-term chemical durability of nuclear waste forms may not be applicable to disposal because the accelerated conditions may not represent disposal conditions. To address this, we examine the corrosion of vitrified archeological materials excavated from the near surface of a ~1500-year old Iron Age Swedish hillfort, Broborg, as an analog for the disposal of vitrified nuclear waste. We compare characterized site samples with corrosion characteristics generated by standard laboratory durability test methods including the product consistency test (PCT), the vapor hydration test (VHT), and the EPA Method 1313 test. Results show that the surficial layer of the Broborg samples resulting from VHT displays some similarities to the morphology of the surficial layer formed over longer timescales in the environment. This work provides improved understanding of long-term glass corrosion behavior in terms of the thickness, morphology, and chemistry of the surficial features that are formed.

On Alteration Rate Renewal Stage of Nuclear Waste Glass Corrosion

MRS Advances ◽  
2020 ◽  
Vol 5 (3-4) ◽  
pp. 111-120 ◽  
Author(s):  
Michael I. Ojovan
Keyword(s):  
Physical State ◽  
Leaching Rate ◽  
Surface Layers ◽  
Near Surface ◽  
High Sodium ◽  
Glass Corrosion ◽  
Initial Stage ◽  
Small Cracks ◽  
Physical Changes

Abstract:The three generically accepted stages of glass corrosion are reviewed with focus on final stage termed alteration rate renewal (or resumption) stage when the glass may re-start corroding with the rate similar to that at the initial stage. It is emphasized that physical state and physical changes that occur in the near-surface layers can readily lead to an effective increase of leaching rate which is similar to alteration rate renewals. Experimental data on long-term (during few decades) corrosion of radioactive borosilicate glass K26 designed to immobilize high-sodium operational NPP radioactive waste evidence on resumption-like effects of radionuclides (137,134Cs) leaching. The cause of that was however related not to chemical changes in the leaching environment but rather to physical state of glass surface due to formation of small cracks and new pristine glass areas in contact with water.

scholarly journals Waste Glass Corrosion Modeling: Comparison with Experimental Results

1993 ◽  
Vol 333 ◽  
Author(s):  
William. L. Bourcier
Keyword(s):  
Borosilicate Glass ◽  
State Theory ◽  
Secondary Phases ◽  
Path Modeling ◽  
Glass Dissolution ◽  
Glass Corrosion ◽  
Controlling Processes ◽  
Rate Limiting ◽  
Geochemical Reaction

ABSTRACTModels for borosilicate glass dissolution must account for the processes of (1) kinetically-controlled network dissolution, (2) precipitation of secondary phases, (3) ion exchange, (4) rate-limiting diffusive transport of silica through a hydrous surface reaction layer, and (5) specific glass surface interactions with dissolved cations and anions. Current long-term corrosion models for borosilicate glass employ a rate equation consistent with transition state theory embodied in a geochemical reaction-path modeling program that calculates aqueous phase speciation and mineral precipitation/dissolution. These models are currently under development. Future experimental and modeling work to better quantify the rate-controlling processes and validate these models are necessary before the models can be used in repository performance assessment calculations.

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