Dissolution Kinetics of UO2. I. Flow-Through tests on UO2.00 Pellets and Polycrystalline Schoepite Samples in Oxygenated, Carbonate/Bicarbonate Buffer Solutions at 25°C

1991 ◽  
Vol 257 ◽  
Author(s):  
Son N. Nguyen ◽  
Homer C. Weed ◽  
Herman R. Leider ◽  
Ray B. Stout
Keyword(s):  
Fission Products ◽  
Dissolution Kinetics ◽  
Spent Fuel ◽  
Bicarbonate Buffer ◽  
Buffer Solutions ◽  
Flow Through ◽  
High Level ◽  
Radionuclide Release ◽  
Kinetics Of

ABSTRACTThe modelling of radionuclide release from waste forms is an important part of the performance assessment of a potential, high-level radioactive waste repository. Since spent fuel consists of UO2 containing actinide elements and other fission products, it is necessary to determine the principal parameters affecting UO2 dissolution and quantify their effects on the dissolution rate before any prediction of long term release rates of radionuclides from the spent fuel can be made.

Dissolution kinetics of meta-torbernite under circum-neutral to alkaline conditions

2009 ◽  
Vol 6 (6) ◽  
pp. 551 ◽  
Author(s):  
Dawn M. Wellman ◽  
Bruce K. McNamara ◽  
Diana H. Bacon ◽  
Elsa A. Cordova ◽  
Ruby M. Ermi ◽  
...  
Keyword(s):  
Dissolution Kinetics ◽  
Coupled Processes ◽  
Effect Of Temperature ◽  
Subsurface Environments ◽  
Alkaline Conditions ◽  
Flow Through ◽  
Uranium Phosphate ◽  
The Stability ◽  
Kinetics Of

Environmental context. Uranium-phosphate minerals have been identified as a long-term controlling phase that limit the mobility of uranium to groundwater in many contaminated subsurface environments. Complex, coupled processes confound the ability to isolate the rates attributed to individual processes. Results of this investigation provide the necessary information to refine current prediction on the release and long-term fate of uranium in subsurface environments. Abstract. The purpose of this investigation was to conduct a series of single-pass flow-through (SPFT) tests to (1) quantify the effect of temperature (23–90°C) and pH (6–10) on meta-torbernite dissolution; (2) compare the dissolution of meta-torbernite to other autunite-group minerals; and (3) evaluate the effect of aqueous phosphate on the dissolution kinetics of meta-torbernite. Results presented here illustrate meta-torbernite dissolution rates increase by ~100× over the pH interval of 6 to 10, irrespective of temperature. The power law coefficient for meta-torbernite, η = 0.59 ± 0.07, is greater than that quantified for Ca-meta-autunite, η = 0.42 ± 0.12. This suggests the stability of meta-torbernite is greater than that of meta-autunite, which is reflected in the predicted stability constants. The rate equation for the dissolution of meta-torbernite as a function of aqueous phosphate concentration is log rdissol (mol m–2 s–1) = –4.7 × 10–13 + 4.1 × 10–10[PO43–].

The Effect of Waste Package Components on Radionuclides Released from Spent Fuel Under Hydrothermal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
Shirley A. Rawson ◽  
William L. Neal ◽  
James R. Burnell
Keyword(s):  
Low Carbon Steel ◽  
Fission Products ◽  
Nuclear Waste Repository ◽  
Low Carbon ◽  
Spent Fuel ◽  
Hydrothermal Conditions ◽  
Waste Package ◽  
Dissolved Species ◽  
Radionuclide Release ◽  
Time Invariant

AbstractThe Basalt Waste Isolation Project has conducted a series of hydrothermal experiments to characterize waste/barrier/rock interactions as a part of its study of the Columbia River basalts as a potential medium for a nuclear waste repository. Hydrothermal tests of 3–15 months duration were performed with light water reactor spent fuel and simulated groundwater, in combination with candidate container materials (low-carbon steel or copper) and/or basalt, in order to evaluate the effect of waste package materials on spent fuel radionuclide release behavior. Solutions were filtered through 400 and 1.8 nm filters to distinguish colloidal from dissolved species. In all experiments, 14C, 129I, and 137Cs occurred only as dissolved species, whereas the actinides occurred in 400 nm filtrates primarily as spent fuel particles. Actinide concentrations in 1.8 nm filtrates were below detection in steel-bearing experiments. In the system spent fuel + copper, apparent time-invariant concentrations of 14C and 137Cs were obtained, but in the spent fuel + steel system, the concentrations of 14C and 137Cs increased gradually throughout the experiments. In experiments containing basalt or steel + basalt, 137Cs concentrations decreased with time. In tests with copper + basalt, 14C and 129I concentrations attained time-invariant values and 137Cs concentrations decreased. Concentrations for the actinides and fission products measured in these experiments were below those calculated from Federal regulations governing radionuclide release.

Durability of potential plutonium wasteforms under repository conditions

2012 ◽  
Vol 76 (8) ◽  
pp. 2911-2918 ◽  
Author(s):  
G. Deissmann ◽  
S. Neumeier ◽  
G. Modolo ◽  
D. Bosbach
Keyword(s):  
Chemical Reactivity ◽  
Final Decision ◽  
Geological Disposal ◽  
Aqueous Environments ◽  
Safety Case ◽  
Key Issues ◽  
The Uk ◽  
High Level ◽  
Radionuclide Release

AbstractSeparated stocks of UK civil plutonium are currently held as a zero value asset in storage, as there is no final decision about whether they should be treated as a resource for future use as nuclear fuel or as waste. Irrespective of future UK government strategies regarding plutonium, at least a portion of the UK civil plutonium inventory will be designated for geological disposal. In this context, we performed a high-level review of the performance of potential wasteforms for the disposal of separated civil plutonium. The key issues considered were the durability and chemical reactivity of the wasteforms in aqueous environments and the long-term radionuclide release under conditions relevant to geological disposal. The major findings of the review, relevant not only to the situation in the UK but to plutonium disposal in general, are summarized in this paper. The review showed that, in the event of a decision being taken to declare plutonium as a waste for disposal, more systematic studies would be required to constrain the wasteform performance under repository conditions in order to derive realistic source terms for a safety case.

scholarly journals The initial dissolution rates of simulated UK Magnox–ThORP blend nuclear waste glass as a function of pH, temperature and waste loading

2015 ◽  
Vol 79 (6) ◽  
pp. 1529-1542 ◽  
Author(s):  
N. Cassingham ◽  
C.L. Corkhill ◽  
D.J. Backhouse ◽  
R.J. Hand ◽  
J.V. Ryan ◽  
...  
Keyword(s):  
Dissolution Kinetics ◽  
Intrinsic Rate ◽  
Waste Glass ◽  
High Level Waste ◽  
Dissolution Mechanism ◽  
Dissolution Rates ◽  
Forward Rates ◽  
Glass Dissolution ◽  
High Level ◽  
Kinetics Of

AbstractThe first comprehensive assessment of the dissolution kinetics of simulant Magnox–ThORP blended UK high-level waste glass, obtained by performing a range of single-pass flow-through experiments, is reported here. Inherent forward rates of glass dissolution were determined over a temperature range of 23 to 70°C and an alkaline pH range of 8.0 to 12.0. Linear regression techniques were applied to the TST kinetic rate law to obtain fundamental parameters necessary to model the dissolution kinetics of UK high-level waste glass (the activation energy (Ea), pH power law coefficient (η) and the intrinsic rate constant (k0)), which is of importance to the post-closure safety case for the geological disposal of vitreous products. The activation energies based on B release ranged from 55 ± 3 to 83 ± 9 kJ mol–1, indicating that Magnox–THORP blend glass dissolution has a surface-controlled mechanism, similar to that of other high-level waste simulant glass compositions such as the French SON68 and LAW in the US. Forward dissolution rates, based on Si, B and Na release, suggested that the dissolution mechanism under dilute conditions, and pH and temperature ranges of this study, was not sensitive to composition as defined by HLW-incorporation rate.

Spent Fuel: Characterization Studies and Dissolution Behaviourunder Disposal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
L. H. Johnson ◽  
D. W. Shoesmith ◽  
S. Stroes-Gascoyne
Keyword(s):  
Release Kinetics ◽  
Fission Products ◽  
Solution Chemistry ◽  
Future Research ◽  
Spent Fuel ◽  
Research Areas ◽  
Dissolution Model ◽  
Underground Disposal ◽  
Radionuclide Release

AbstractThe concept of disposal of unreprocessed spent fuel has now been under study internationally for over ten years. Considerable progress has been made in understanding the factors that will control radionuclide release from spent fuel in an underground disposal vault. This progress is reviewed and the research areas of significance in providing further data for source term models are discussed. Key areas for future research are identified; these include improved characterization of spent fuel to determine the inventories of fission products at grain boundaries, together with their release kinetics; and a better understanding of the effects of solution chemistry on spent fuel dissolution, in particular the effects of salinity, redox chemistry, and radiolysis of groundwater. Approaches to modelling the dissolution of spent fuel are discussed, and a possible approach for developing an oxidative dissolution model is outlined.

Constraints by Experimental Data for Modeling of Radionuclide Release from Spent Fuel

1989 ◽  
Vol 176 ◽  
Author(s):  
Bernd Grambow ◽  
L.O. Werme ◽  
R.S. Forsyth ◽  
J. Bruno
Keyword(s):  
Experimental Data ◽  
Product Layer ◽  
Spent Fuel ◽  
Upper Limits ◽  
The Matrix ◽  
The Usa ◽  
Oxidant Production ◽  
Release Data ◽  
Radionuclide Release

ABSTRACTComparison of spent fuel corrosion data from nuclear waste management projects in Canada, Sweden and the USA strongly suggests that the release of 90Sr to the leachant can be used as a measure of the degradation (oxidation/dissolution) of the fuel matrix. A surprisingly quantitative similarity in the 90 Sr release data for fuel of various types (BWR, PWR, Candu), linear power ratings and burnups leached under oxic conditions was observed in the comparison. After 1000 days of leachant contact, static or sequential, the fractional release rates for 90Sr (and for cesium nuclides) were of the order of 10−7/d.The rate of spent fuel degradation (alteration) under oxic conditions can be considered to be controlled either by the growth rates of secondary alteration products, by oxygen diffusion through a product layer, by the rate of formation of radiolytic oxidants or by solubility-controlled dissolution of the matrix. These processes are discussed. Methods for determining upper limits for long-term 90Sr release, and hence fuel degradation, have been derived from the experimental data and consideration of radiolytic oxidant production.

What really happens to the “vanishing overpack”?

2012 ◽  
Vol 1475 ◽  
Author(s):  
I. G. McKinley ◽  
F. B. Neall ◽  
E. M. Scourse ◽  
H. Kawamura
Keyword(s):  
Radionuclide Transport ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Failure Times ◽  
Bentonite Buffer ◽  
Physical Barriers ◽  
Safety Assessments ◽  
High Level ◽  
Radionuclide Release

ABSTRACTConcepts for the disposal of high-level radioactive waste (HLW) and spent fuel (SF) in several countries include a massive steel overpack within a bentonite buffer. In past conservative safety assessments to demonstrate feasibility of geological disposal, overpacks are assumed to provide complete containment for a given lifetime, after which all fail simultaneously. After failure, they are ignored as physical barriers to radionuclide transport. In order to compare different repository designs for specific sites, however, a more realistic treatment of overpack failure and its subsequent behaviour is needed. In addition to arguing for much longer lifetimes before mechanical failure and a distribution of overpack failure times, such assessment indicates that the presence of the failed overpack greatly constrains radionuclide release from the waste matrix and subsequent migration through the engineered barrier system. It also emphasises the key role of the bentonite buffer and the need to be able to assure its performance over relevant timescales.

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.

Use of Level 1 PSA in the Frame of EPR Licensing

2014 ◽  
Author(s):  
Gabriel Georgescu ◽  
Patricia Dupuy ◽  
Francois Corenwinder
Keyword(s):  
Probabilistic Methods ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Spent Fuel Pool ◽  
Level 1 ◽  
Water Reactors ◽  
High Level ◽  
Design Construction

The French “Technical Guidelines for the design and construction of the next generation of NPPs with Pressurized Water Reactors” specify that the safety demonstration has to be achieved in a deterministic way, supplemented by probabilistic methods. In this context, for the EPR reactor of Flamanville (EPR-FA3), the PSA has been used from the beginning of the design. In the frame of the application for commissioning of EPR-FA3, EdF has to provide an “as-build” full scope PSA for the reactor and for the spent fuel pool, covering the internal events, as well as the internal and external hazards of significant impact. Some of these probabilistic studies were developed and evolved during the EPR design (PSA for “internal events”, specific studies for practically eliminated sequences, long term accident sequences…) and were analysed by IRSN, as French Safety Authority (ASN) technical support, at different EPR project stages (initial design, detailed design, construction application….) leading to many design and studies improvements. Today, in order to make the analysis of the application for the commissioning of the EPR-FA3 reactor more effective, the “anticipated” analysis of this application is in progress in France. In this context, the updated versions of the level 1 probabilistic studies for internal events and hazards were analysed by IRSN in 2013. The results and conclusions of this analysis were presented by IRSN early 2014 during a dedicated meeting of French Standing Group of experts for Reactors safety (SGR). The paper presents the analysis performed by IRSN of EdF EPR-FA3 level 1 probabilistic studies, highlighting the role of PSA to the achievement of high level of safety of EPR reactor.

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