Effect of Spent Fuel Burnup and Composition on Alteration of the U(Pu)O2 Matrix

2000 ◽  
Vol 663 ◽  
Author(s):  
C. Jégou ◽  
S. Peuget ◽  
J.F. Lucchini ◽  
C. Corbel ◽  
V. Broudic ◽  
...  
Keyword(s):  
Release Kinetics ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Static Mode ◽  
Mox Fuel ◽  
The Matrix ◽  
Leaching Experiments ◽  
Experimental Findings ◽  
Radionuclide Release ◽  
Kinetics Of

ABSTRACTFor a potential performance assessment of direct disposal of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified.Leaching experiments were performed with five spent fuel samples from French power reactors (four UO2 fuel samples with burnup ratings of 22, 37, 47 and 60 GWd·tHM−1 and a MOX fuel sample irradiated to 47 GWd·tHM−1) to determine the release kinetics of the matrix containing most (over 95%) of the radionuclides. The experiments were carried out with granitic groundwater on previously leached sections of clad fuel rods in static mode, in an aerated medium at room temperature (25°C) in a hot cell.After 1000 or 2000 days of leaching, the Sr/U congruence ratios for all the UO2 fuel samples ranged from 1 to 2; allowing for the experimental uncertainty, strontium can thus be considered as a satisfactory matrix alteration tracer. No significant burnup effect was observed on the alteration of the UO2 fuel matrix. The daily strontium release factor was approximately 1 à 10−7 d−1 for UO2 fuel, and five to six times higher for MOX fuel. Several alteration mechanisms (radiolysis, solubility, precipitation/clogging) are examined to account for the experimental findings.

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.

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.

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.

Evaluation of Kollicoat SR 30D and Kollicoat EMM 30D as Matrix Former for Controlled Release Drug Delivery

1970 ◽  
Vol 4 (1) ◽  
Author(s):  
Mohammad Salim Hossain ◽  
Reza-ul Jalil ◽  
Selim Reza ◽  
Mohiuddin Abdul Quadir ◽  
CF Hossain
Keyword(s):  
Controlled Release ◽  
Drug Release ◽  
Matrix Effects ◽  
Release Kinetics ◽  
Polymeric Systems ◽  
Release Drug ◽  
The Matrix ◽  
Kinetics Of ◽  
Order Pattern ◽  
Drug Liberation

Efficiency of kollicoat EMM 30 D and SR 30D as matrix forming material was investigated. It was found that, theophylline loaded granules prepared with these two polymers could not sustain drug release for a significant period of time. However, compression of these granules into tablets retarded drug release for up to 8 hours. Release was faster from EMM 30D polymeric system than that from SR 30D matrix. Effects of fillers and rate modifiers on drug liberation have been assessed. Incorporation of Avicel RC 591 and starch caused substantial release of theophylline from both the polymeric systems. Avicel PH 101 intensified the retardation effect of both EMM 30D and SR 30D on theophylline release. HPMC 50 cps, when added to the matrix, caused the release of theophylline to follow near zero order pattern. Increasing the content of HPMC in both EMM 30D and SR 30D compressed tablets decreased the rate and extent of theophylline release. In the presence of excipients, no significant differences between rate and extent of drug release from EMM 30D and SR 30D systems were found. Biexponential equation was applied to explore and explain drug release kinetics. It was found that drug release followed Fickian or case I kinetics from EMM 30D compressed tablet while anomalous or non-fickian kinetics of drug release was observed for SR 30D system. Key words: Kolliocoat SR 30D, Kollicoat EMM 30D, Theophylline, Matrix system, Controlled release Dhaka Univ. J. Pharm. Sci. Vol.4(1) 2005 The full text is of this article is available at the Dhaka Univ. J. Pharm. Sci. website

Unsaturated Zone Waters From the Nopal I Natural Analog, Chihuahua, Mexico - Implications for Radionuclide Mobility at Yucca Mountain

1999 ◽  
Vol 556 ◽  
Author(s):  
David A. Pickett ◽  
William M. Murphy
Keyword(s):  
Unsaturated Zone ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Path Modeling ◽  
Rock Interaction ◽  
Colloidal Transport ◽  
Natural Analog ◽  
Initial Silica ◽  
Radionuclide Release

AbstractChemical and U-Th isotopic data on unsaturated zone waters from the Nopal I natural analog reveal effects of water-rock interaction and help constrain models of radionuclide release and transport at the site and, by analogy, at the proposed nuclear waste repository at Yucca Mountain. Geochemical reaction-path modeling indicates that, under oxidizing conditions, dissolution of uraninite (spent fuel analog) by these waters will lead to eventual schoepite precipitation regardless of initial silica concentration provided that groundwater is not continuously replenished. Thus, less soluble uranyl silicates may not dominate the initial alteration assemblage and keep dissolved U concentrations low. Uranium-series activity ratios are consistent with models of U transport at the site and display varying degrees of leaching versus recoil mobilization. Thorium concentrations may reflect the importance of colloidal transport of low-solubility radionuclides in the unsaturated zone.

scholarly journals New in vitro model to evaluate kinetics of antimycobacterial drug release from bioresorbable polymeric carriers

2020 ◽  
pp. 10-15
Author(s):  
SN Andreevskaya ◽  
TG Smirnova ◽  
EN Antonov ◽  
LN Chernousova ◽  
SE Bogorodsky ◽  
...  
Keyword(s):  
Drug Release ◽  
Culture Medium ◽  
Culture Media ◽  
Release Kinetics ◽  
Prolonged Release ◽  
Wide Range ◽  
Polymeric Carriers ◽  
The Matrix ◽  
Kinetics Of

Sustained-release drugs against tuberculosis are a promising approach to therapy since they positively affect patient compliance with long regimens, especially when it comes to the multidrug-resistant form of the disease. Conventional UV-visible spectroscopy does not work well with multicomponential culture media used for growing M. tuberculosis. The aim of this study was to develop a method for evaluating the kinetics of anti-tuberculosis drug released from bioresorbable polymeric carriers suitable for screening a wide range of encapsulated prolonged-release drugs and identifying the best performing candidate. While studying the growth dynamics of the laboratory susceptible strain M. tuberculosis H37Rv in the presence of different levofloxacin concentrations (from 0.03 to 0.4 μg/ml), we developed a model, which is essentially a set of 2 parallel experiments evaluating the kinetics of drug release into the culture medium. The results of these 2 experiments conducted on 3 encapsulated forms of levofloxacin loaded onto bioresorbable polymeric PLGA carriers (particles sized 50 μm and 100 μm and the matrix) revealed that release kinetics of the drug largely depended on the type of polymeric carrier. The best encapsulation of the antibiotic and its gradual release into the culture medium was observed for the matrix. All experiments were run in 3 replicates. The obtained data were analyzed using descriptive statistics.

Rim Instant Release Radionuclide Inventory From French High Burnup Spent UOX Fuel

2009 ◽  
Vol 1193 ◽  
Author(s):  
D. Roudil ◽  
C. Jegou ◽  
V. Broudic ◽  
M. Tribet
Keyword(s):  
Small Volume ◽  
Source Term ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Waste Container ◽  
High Burnup ◽  
Radionuclide Release ◽  
Long Term Behavior ◽  
The Impact

AbstractTo assess the long-term behavior of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified. The instant release fraction (IRF) source term assumed to be instantaneously accessible to water after the failure of the waste container. Some IRF values for different kinds of spent fuel are available in the literature. However, the possible contribution the rim restructured zone for high-burnup UOX fuels, was not necessary taken into account. A specific study of the leaching behavior of the rim zone has been carried out on a UOX fuel sample with a burnup of 60 GWd·t-1 and 2.8% FGR. The 134/137Cs and 90Sr rim IRF are effectively higher than the gap and grain boundary inventories (respectively 4.4 wt% and 0.3 wt% of the total Cs and Sr inventories). Nevertheless because of the relative small volume of this zone in the pellet, the impact of the rim inventory appears to be limited and the complete Cs and Sr IRF (gap, grain boundaries and rim), were estimated at 1.85 and 0.3 wt%, respectively

The Kinetic of Dissolution of Irradiated UO2 Fuel Under Oxidising Conditions in Flow Experiments

2001 ◽  
Author(s):  
J. Cobos ◽  
J. A. Serrano ◽  
J. P. Glatz ◽  
B. Sätmark-Christiansen ◽  
J. de Pablo
Keyword(s):  
Redox Potential ◽  
Room Temperature ◽  
Dissolution Kinetics ◽  
Deionized Water ◽  
Matrix Phase ◽  
Spent Fuel ◽  
Carbonate Concentration ◽  
Leaching Experiments ◽  
Kinetics Of ◽  
Flow Experiments

Abstract These leaching experiments report the effects of four important parameters (redox potential, pH, carbonate concentration and temperature) on the dissolution kinetics of the spent fuel matrix phase. The kinetic of dissolution of irradiated UO2 fuel has been studied in deionized water and synthetic granite groundwater under oxidising conditions at room temperature.

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.

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