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.

Leaching Study of the Behaviour of Spent Fuel and SIMFUEL Under Simulated Granitic Repository Conditions

2001 ◽  
Author(s):  
J. A. Serrano ◽  
J. Quiñones ◽  
J. Cobos ◽  
P. Diaz Arocas ◽  
V. V. Rondinella ◽  
...  
Keyword(s):  
Fission Products ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Experimental Conditions ◽  
Radiation Fields ◽  
Leaching Experiments ◽  
Irradiated Fuel ◽  
Fractional Release ◽  
Intense Radiation ◽  
Chemical Analogue

Abstract The leaching behaviour of spent fuel is of importance for the concept of direct storage of spent fuel. The aim of this study was to study UO2 irradiated fuel under simulated granitic repository conditions. In parallel with these spent fuel tests, SIMFUEL leaching studies were also performed. Direct comparisons between spent fuel and its chemical analogues, SIMFUEL, are often difficult. On one hand, because of the differences existing between spent fuel and SIMFUEL. E.g., for irradiated fuel: different origin and burnup, presence of intense radiation fields, hence radiolysis effects, or formation of cracks and pores due to the volatile fission products, hence larger surface area. On the other hand, because of different experimental procedures used by different authors. This work presents results of sequential leaching experiments in synthetic granite water in equilibrium with a cylinder of granite at room temperature in air using spent UO2 fuel and SIMFUEL. The experimental conditions and procedure for irradiated and non-irradiated materials were kept similar as much as possible. The specimens used were UO2 (43 MWd/kgU) and SIMFUEL (simulating a burnup of 30 MWd/kgU) as non-irradiated chemical analogue. A thermodynamic study by means of geochemistry codes was also performed. Differences both in fractional release and in uranium concentration in the leachate were found. The highest fractional release of uranium was measured for UO2 spent fuel. Candidate solid phases calculated for controlling the uranium solubility were soddyite ((UO2)2(SiO4)·2H2O) in the case of spent fuel and haiweete (Ca(UO2)2(Si2O5)3·5H2O) for SIMFUEL. Further work is ongoing to characterise the surfaces of the leached fuel samples and to try to confirm the preliminary attempts to identify reprecipitated secondary phases. Comparison of some fission product release between spent fuel and SIMFUEL was also performed.

Coprecipitation Experiments Using Simulated Spent Fuel Solutions in the Presence of Metallic Iron in Synthetic Bentonitic-Granitic Water Under Oxidising Conditions

2004 ◽  
Vol 824 ◽  
Author(s):  
Javier Quiñones ◽  
Ángel González de la Huebra ◽  
Aurora Martínez Esparza
Keyword(s):  
Metallic Iron ◽  
Room Temperature ◽  
Uranium Concentration ◽  
Solid Phase ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Bulk Solid ◽  
Show Evidence ◽  
Ph Range ◽  
Iron Material

AbstractThis paper presents the results obtained from coprecipitation experiments of uranyl solutions in the presence of metallic iron and/or its alteration phases in synthetic bentonitic-granitic composition water. Experiments were done under oxidising conditions at room temperature. The pH range covered was 7.4 – 8.8. Changes in the uranium concentrations and the characterisation of the secondary phases formed in the experiments were done using XRD and SEM-EDS and are presented herein.Final uranium concentration values were in the range of 2·10−5 – 5·10−4 mol (kg of H2O)−1. In all cases, results from these experiments did not show evidence of a clear effect due to the presence of iron (metallic or previously corroded) on the uranium concentration. These data were similar to those obtained by coprecipitation in similar conditions but in absence of iron material. Boltwoodite was observed [K2(UO2)2(SiO3)2(OH)2·3H2O] in iron surface materials and characterized in all experiments. Based on the experimental data obtained (uranium concentration in solution and bulk solid phase characterisation) we propose the following surface-mediated reaction:2 K+ + 2 H4SiO4 + 2 UO2+2 + 3 H2O ⇔ K2(UO2)2(SiO3)2(OH)2·3H2O + 6 H+

Release of U, REE and Th From Palmottu Granite

2008 ◽  
Vol 1107 ◽  
Author(s):  
Mira K. Markovaara-Koivisto ◽  
Nuria Marcos ◽  
David Read ◽  
Antero Lindberg ◽  
Marja SiitariKauppi ◽  
...  
Keyword(s):  
Metal Mobility ◽  
Carbonate Content ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Mineralogical Characterization ◽  
Crushed Material ◽  
Icp Ms ◽  
Before And After ◽  
High Carbonate ◽  
Experimental Protocols

AbstractInterpretation of trace metal mobility in geological environments is often hampered by conflicting data from alternative experimental protocols and the lack of detailed mineralogical characterization of the host medium. To illustrate this issue, the release of uranium, thorium and the rare earth elements (REE) was investigated in polished rock slab samples from the U-Th deposit at the Palmottu Natural Analogue study site (SW Finland) by means of leaching experiments. The samples were sequentially leached with artificial groundwater of moderately high carbonate content at pH8, and nitric acid solutions at pH5 and pH3. The mineralogy and composition of the U, Th and REE mineral phases was studied using SEM-WDS and EDAX methods before and after each leaching step. In parallel, leaching was carried out on crushed material of the same samples and the leachates analysed by ICP-MS.The most notable U minerals are uraninite, uranophane and two secondary U-Pb phases. Thorium occurs predominantly in monazite and at lower concentrations in uraninite. Accessory thorite is also present, which together with monazite contains most of the REE. Differential leaching of the elements was noted across all phases on the timescale of the experiments. Uraninite is partly dissolved at pH3. The main secondary uranium phase, uranophane, was stable in moderately acidic solution, but easily dissolved in the artificial groundwater and at pH3. Some release of REE was observed although the main REE-bearing phase, monazite, showed no evidence of degradation.This study provides insights in the preferential release of radionuclides in granitic bedrock. An understanding of these processes is essential when assessing the safety of a spent fuel repository. Once released from the primary waste form U is expected to precipitate as secondary phases within micro fractures, as observed at Palmottu and numerous other deposits.

Source Term Assessment: The ARIANE Programme

2001 ◽  
Author(s):  
Marc Lippens ◽  
Danièle Boulanger ◽  
Jacques Basselier ◽  
Stefaan Van Winckel ◽  
Beat Wernli ◽  
...  
Keyword(s):  
Source Term ◽  
Fission Products ◽  
Large Set ◽  
Highly Qualified ◽  
Icp Ms ◽  
Measurement Results ◽  
Activation Products ◽  
Extended Analysis ◽  
Fission And Activation Products ◽  
Irradiated Fuel

Abstract The prediction of the source term of actinides, fission and activation products is of major importance in numerous nuclear fields. An experimental programme of radiochemistry (called ARIANE) was carried out on MOX and UO2 fuels irradiated in PWR and BWR conditions, reaching burnups of 35 up to 55 GWd/tM. Besides the providing of a large set of experimental data on the irradiated fuel inventory, a second objective of the programme was to confer on these experimental results reliable and minimised uncertainties. About 50 isotopes of actinides and fission products were selected. Their contents were measured using techniques as TIMS, ICP-MS, alpha, beta and gamma spectrometry. The measurement results were crosschecked by three highly qualified laboratories and recommended values were provided after extended analysis and confrontation of the results. For most of the measured isotopes, the target uncertainties were reached and even reduced, providing a unique database for irradiated fuel characterisation and for code qualification.

Characterization of Different Types of Biomass Wastes Using Thermogravimetric and ICP-MS Analyses

2020 ◽  
Vol 70 (12) ◽  
pp. 4594-4600
Keyword(s):  
Thermogravimetric Analysis ◽  
Food Industry ◽  
Wood Waste ◽  
Biomass Ash ◽  
Minor Elements ◽  
Ms Analysis ◽  
Icp Ms ◽  
Different Types

The purpose of this study was to characterize some types of biomass wastes resulted from different activities such as: agriculture, forestry and food industry using thermogravimetric and ICP-MS analyses. Also, it was optimized an ICP-MS method for the determination of As, Cd and Pb from biomass ash samples. The ICP-MS analysis revealed that the highest concentration of metals (As, Cd, Pb) was recorded in the wood waste ash sample, also the thermogravimetric analysis indicated that the highest amount of ash was obtained for the same sample (26.82%). The biomass wastes mentioned in this study are alternative recyclable materials, reusable as pellets and briquettes. Keywords: ash, biomass, ICP-MS, minor elements, TG

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.

scholarly journals Trace and minor elements in galena: A reconnaissance LA-ICP-MS study

2015 ◽  
Vol 100 (2-3) ◽  
pp. 548-569 ◽  
Author(s):  
L. George ◽  
N. J. Cook ◽  
C. L. Ciobanu ◽  
B. P. Wade
Keyword(s):  
Minor Elements ◽  
Icp Ms ◽  
Trace And Minor Elements

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).

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.

Sign in / Sign up

Export Citation Format

Share Document