Processes Controlling Radionuclide Release from Spent Fuel

1994 ◽  
Vol 353 ◽  
Author(s):  
A. Loida ◽  
B. Grambow ◽  
H. Geckeis ◽  
P. Dressler
Keyword(s):  
Volume Ratio ◽  
Reaction Rates ◽  
Geometric Constraints ◽  
Spent Fuel ◽  
Oxidative Dissolution ◽  
Strong Decrease ◽  
Radiolysis Products ◽  
Upper Limits ◽  
Fuel Behavior ◽  
Radionuclide Release

AbstractDissolution of spent fuel has been studied in saline, anaerobe, carbonate free solutions. Processes controlling spent fuel dissolution and associated radionuclide release are radiolytically controlled oxidative dissolution, sorption on container, solubility and coprecipitation. Upper limits for oxidative dissolution rates are given by the production rates of oxidative radiolysis products. This limitation leads to a strong decrease in surface area normalized reaction rates with increasing surface to volume ratio (S/V) and imposes geometric constraints on prediction of spent fuel behavior in a repository. Solution concentrations of Am during spent fuel corrosion were about 5 orders of magnitude lower than the solubility of Am(OH)3(s) and are likely controlled by coprecipitation. Pu concentrations may be controlled by Pu(VI) or Pu(IV) (hydr)oxides.

Chemical Durability of High Burnup Lwr-Spent Fuel in Concentrated Salt Solutions

1993 ◽  
Vol 333 ◽  
Author(s):  
A. Loida ◽  
B. Grambow ◽  
P. Dressier ◽  
K. Friese ◽  
H. Geckeis ◽  
...  
Keyword(s):  
Surface Area ◽  
Volume Ratio ◽  
Sample Surface ◽  
Solution Concentration ◽  
Reaction Rates ◽  
Spent Fuel ◽  
Intrinsic Dissolution ◽  
Fine Grained ◽  
High Burnup ◽  
Saturation Effects

ABSTRACTHigh-burnup (<50 MWd/kgU) spent fuel samples of various sizes were exposed to NaCl solutions under static, anaerobic and reducing conditions. The accumulated corrosion time was about 200 days. Gas phase and leach solutions were analyzed. By dissolving mm sized fragments in large volumes of solution, saturation effects were avoided and upper limits for intrinsic dissolution rates of about 5-20 mg/(m2d) were measured. Surface area normalized reaction rates were significantly lower when using fine grained fuel powder (estimated sample surface area to solution volume ratio S/V ca. 3000 m-1), indicating saturation effects. The maximum concentrations of Pu and Am in the tests are close to reported solubility limited concentrations in pure 5m NaCl solutions in the absence of radiolysis effects. The presence of iron effectively reduces the solution concentration of all measured radionuclides (except Cs).

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.

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.

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).

Modelling the Release of Radionuclides from the Spent Fuel and their Transport in the Near Field: a Coupled Approach

2003 ◽  
Vol 807 ◽  
Author(s):  
Juan Merino ◽  
Esther Cera ◽  
Jordi Bruno ◽  
Aurora Martínez-Esparza
Keyword(s):  
Near Field ◽  
Radioactive Decay ◽  
Secondary Phase ◽  
Coupled System ◽  
Integrated Approach ◽  
Integrated Systems ◽  
Advective Transport ◽  
Spent Fuel ◽  
Oxidative Dissolution ◽  
Phase Precipitation

ABSTRACTIn this work we have developed a model for the release of radionuclides from the spent fuel coupled with their transport through the near field. A compartmental approach has been used, as this methodology is well suited to model integrated systems. Several processes have been taken into account: oxidative dissolution of the spent fuel matrix, radioactive decay and chains, diffusive and advective transport, retardation by sorption and secondary phase precipitation. Results illustrate the complex evolution of the radionuclide concentrations in the gap and the near field. Hence, the main conclusion from this study is the requirement to model this coupled system using a compartmental integrated approach.

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.

Modelling Oxidative Dissolution of Spent Fuel

1996 ◽  
Vol 465 ◽  
Author(s):  
Ivars Neretnieks
Keyword(s):  
Redox Reaction ◽  
Uranium Oxide ◽  
Reaction Rate ◽  
Spent Nuclear Fuel ◽  
Large Fraction ◽  
Reaction Rates ◽  
Spent Fuel ◽  
First Case ◽  
Rate Of Movement

ABSTRACTSpent nuclear fuel will, by the radiation, split nearby water into oxidizing and reducing compounds. The reducing compounds are mostly hydrogen that will diffuse away. The remaining oxidizing compounds can oxidize the uranium oxide of the fuel and make it more soluble. The oxidised uranium will dissolve and diffuse away. The nuclides previously incorporated in the spent fuel matrix can then be released and also migrate away from the fuel.A model is proposed where the produced oxidizing species compete for reaction with the fuel and for escaping out of the system. The chemical reaction rate of oxygen and fuel is taken from literature values based on experiments. The escape rate of oxidants to a receding redox front in the backfill is modelled assuming a redox reaction of oxidizing component and reducing component in the surrounding. The rate of movement of the redox front is determined from the rate of production of oxidants. This is estimated using a previously devised model that has been calibrated to in situ observed radiolysis.Three cases are modelled. In the first case it is assumed that the reducing compound is insoluble and that the reaction between oxygen and reducing mineral is very fast. In the second case it is assumed that the reducing component has a known solubility and that it can migrate to meet the oxygen and quickly react. In a third case a finite reaction rate is modelled between the oxygen and the reducing species.The sample calculations show that if the reducing mineral has to be supplied from the backfill a large fraction of the spent fuel could be oxidised. If the corrosion products of a degraded steel canister can supply the reducing species and the redox reaction is fast, very small amounts of the fuel could be oxidised. Literature data indicate that the redox reaction rate may not be so fast that it can be considered instantaneous and then a considerable fraction of the fuel could be oxidised. The model gives a means of exploring which mechanisms and data may be of most importance for radiolytic fuel dissolution, but the realism of the data and the model must be tested further. There is a lack of understanding and data on reaction rates, heterogeneous as well as homogeneous. This is crucial to the results.

Coprecipitation Studies from Simfuel Solutions in 5 m NaCl Media

1996 ◽  
Vol 465 ◽  
Author(s):  
P. Diaz-Arocas ◽  
J. Garcia-Serrano
Keyword(s):  
Fission Products ◽  
Solution Concentration ◽  
Laser Fluorescence ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Minor Elements ◽  
Icp Ms ◽  
Upper Limits ◽  
Ph Range ◽  
Irradiated Fuel

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.

Sign in / Sign up

Export Citation Format

Share Document