Aqueous leaching of ADOPT and standard UO2 spent nuclear fuel under H2 atmosphere

MRS Advances ◽  
2020 ◽  
Vol 5 (3-4) ◽  
pp. 167-175
Author(s):  
Alexandre Barreiro Fidalgo ◽  
Olivia Roth ◽  
Anders Puranen ◽  
Lena Z. Evins ◽  
Kastriot Spahiu
Keyword(s):  
Gas Phase ◽  
Nuclear Fuel ◽  
Preliminary Data ◽  
Spent Nuclear Fuel ◽  
Uranium Concentration ◽  
Hydrogen Atmosphere ◽  
Dissolution Behavior ◽  
Leaching Behavior ◽  
New Type ◽  
Failure Scenario

ABSTRACTLeaching results to compare the dissolution behavior of a new type of fuel with additives (Advanced Doped Pellet Technology, ADOPT) with standard UO2 fuel are presented. Both fuels were irradiated in the same assembly of a commercial boiling water reactor to a local burnup of ∼58 MWd/kgU. Fuel fragments are leached in simplified groundwater in two autoclaves under hydrogen atmosphere, representing conditions in a canister failure scenario resulting in water intrusion for a spent nuclear fuel repository. Preliminary results indicate the uranium concentration decreased to 3-4x10-8 M after 421 days, slightly above the solubility of amorphous UO2. Xe has been detected in the gas phase of both autoclaves. The concentration of Cs and I seems to gradually approach constant values, yet the redox sensitive elements continue to slowly increase with time. The preliminary data obtained supports the hypothesis that there is no major difference in leaching behavior between the two fuels.

Modelling the Activation of H2 on Spent Fuel Surface and Inhibiting Effect of UO2 Dissolution

2013 ◽  
Vol 1518 ◽  
pp. 133-138 ◽  
Author(s):  
L. Duro ◽  
O. Riba ◽  
A. Martínez-Esparza ◽  
J. Bruno
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Pellet ◽  
Dissolution Behavior ◽  
Dissolution Mechanism ◽  
Water Radiolysis ◽  
Spent Fuel ◽  
Secondary Phases ◽  
The Matrix ◽  
Fuel Storage

ABSTRACTThe dissolution of spent nuclear fuel is defined in two different time steps, i) the Instant Release Fraction (IRF) occurring shortly after water contacts the solid spent fuel and responsible of the fast release of those radionuclides that have been accumulated in the zones of the spent fuel pellet with low confinement, such as gap and grain boundaries and ii) the long term release of radionuclides confined in the spent fuel matrix, much slower and dependent on the conditions of the water that contacts the spent fuel.Several models have been developed to date to explain the dissolution behavior of spent nuclear fuel under disposal conditions. The Matrix Alteration Model (MAM) is one of the most evolved radiolytic models describing the dissolution mechanism in which an Alteration/Dissolution source term model is based on the oxidative dissolution of spent fuel. Under deep repository conditions and at the expected of water contacting time (after 1000 years of spent fuel storage), α radiation will be the main contributor to water radiolysis. In the current study, simulations evaluating the effect of surface area on the alteration/dissolution of spent fuel matrix are performed considering different particle sizes of spent fuel and simulations integrating the actinides dissolution have been performed considering the precipitation of secondary phases.

Studies of sorption of Th and Np on UO2 and TiO2

2000 ◽  
Vol 663 ◽  
Author(s):  
Yngve Albinsson ◽  
Hans Nilsson ◽  
Anna-Maria Jakobsson
Keyword(s):  
Distribution Coefficient ◽  
Surface Area ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
No Reduction ◽  
Hydrogen Atmosphere ◽  
High Extent ◽  
The Moment ◽  
Tetravalent State ◽  
Pentavalent State

ABSTRACTWhen designing an underground repository for spent nuclear fuel it is important to know if and to what extent Np and Pu are reduced. At the moment only Np has been studied, but further investigations with Pu are planned.Np in the pentavalent state and Th in the tetravalent state (234Th, about 10−9 M) have been used to study the sorption onto TiO2 (no reduction is expected) and onto UO2 with different atmospheres (air, nitrogen and nitrogen and hydrogen at 5 MPa pressure), and with different Np- concentrations (about 10−12M, 239Np and 10−6 M,237Np).Th is sorbed to a high extent at pH>3 on both TiO2 and UO2. The sorption increases 3 orders of magnitude over one pH-unit when log Ka (the distribution coefficient with respect to surface area) is plotted against pH.The Np sorption on TiO2 is independent of the Np concentration and shows a slope of one when pH is plotted against log Ka. The higher sorption of Np onto UO2 compared with TiO2 indicates that Np is partially reduced even when it is in contact with air. No large differences can be observed for Np between the nitrogen and hydrogen-atmosphere, indicating that the absence of oxygen is the important factor for the reduction. The different Np concentrations had no great impact on the reduction/sorption on UO2.

Study of the release of the fission gases (Xe and Kr) and the fission products (Cs and I) under anoxic conditions in bicarbonate water

2015 ◽  
Vol 1744 ◽  
pp. 35-41 ◽  
Author(s):  
Ernesto González-Robles ◽  
Markus Fuß ◽  
Elke Bohnert ◽  
Nikolaus Müller ◽  
Michel Herm ◽  
...  
Keyword(s):  
Gas Phase ◽  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Collaborative Project ◽  
Fuel Rod ◽  
Release Ratio

ABSTRACTFor safety assessment analyses of the disposal of spent nuclear fuel (SNF) in deep geological repositories it is indispensable to evaluate the contribution of fission products to the instant release fraction (IRF). During the last three years the EURATOM FP7 Collaborative Project, “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)” was carried out to get a better understanding of the IRF.Within CP FIRST-Nuclides, a leaching experiment with a cladded SNF pellet was performed in bicarbonate water (19 mM NaCl + 1 mM NaHCO3) under Ar /H2 atmosphere over 333 days. The cladded SNF pellet was obtained from a fuel rod segment which was irradiated in the Gösgen pressurized water reactor; the average burn-up of the segment was 50.4 MWd/kgUO2. In the multi-sampling experiment, gaseous and liquid samples were taken periodically. The moles of the fission gases Kr and Xe released in the gas phase and those of 129I and 137Cs released in solution were measured. Cumulative release fractions of (1.6 ± 0.2)·10-1 fission gases, (1.6 ± 0.1)·10-1129I and (3.9 ± 0.2)·10-2 137Cs, respectively, were achieved after 333 days of leaching. Accordingly the release ratio of fission gases to 129I was 1:1 and the release ratio of fission gases to 137Cs was 4:1, respectively.

Materials Characteristics and Dissolution Behavior of Spent Nuclear Fuel

MRS Bulletin ◽  
1994 ◽  
Vol 19 (12) ◽  
pp. 24-27 ◽  
Author(s):  
L.H. Johnson ◽  
L.O. Werme
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Waste Form ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Power Stations ◽  
Geologic Disposal

The geologic disposal of spent nuclear fuel is currently under consideration in many countries. Most of this fuel is in the form of assemblies of zirconium-alloy-clad rods containing enriched (1–4% 235U) or natural (0.71% 235U) uranium oxide pellets. Approximately 135,000 Mg are presently in temporary storage facilities throughout the world in nations with commercial nuclear power stations.Safe geologic disposal of nuclear waste could be achieved using a combination of a natural barrier (the host rock of the repository) and engineered barriers, which would include a low-solubility waste form, long-lived containers, and clay- and cement-based barriers surrounding the waste containers and sealing the excavations.A requirement in evaluating the safety of disposal of nuclear waste is a knowledge of the kinetics and mechanism of dissolution of the waste form in groundwater and the solubility of the waste form constituents. In the case of spent nuclear fuel, this means developing an understanding of fuel microstructure, its impact on release of contained fission products, and the dissolution behavior of spent fuel and of UO2, the principal constituent of the fuel.

Spent Fuel Leaching in the Presence of Corroding Iron

MRS Advances ◽  
2017 ◽  
Vol 2 (12) ◽  
pp. 681-686
Author(s):  
Anders Puranen ◽  
Alexandre Barreiro ◽  
Lena Z. Evins ◽  
Kastriot Spahiu
Keyword(s):  
Gas Phase ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Hydrogen Effect ◽  
Spent Fuel ◽  
Iron Corrosion ◽  
Reducing Conditions ◽  
Partial Pressures ◽  
Corrosion Of Iron ◽  
Copper Shell

ABSTRACTThe Swedish spent nuclear fuel canister design KBS-3 consists of a cylindrical copper shell surrounding an iron insert that holds the spent fuel. Like in most other canister designs the mass of iron constitutes the majority of the canister weight. In order for groundwater to access the spent fuel in a future repository the copper shell must fail and iron corrosion occur. Spent nuclear fuel dissolution will therefor likely proceed under conditions of simultaneous anoxic iron corrosion. The iron corrosion can likely suppress the spent fuel release by creation of strongly reducing conditions from Fe(II) formation and the generation of large quantities of hydrogen. Redox sensitive radionuclides may either be reductively precipitated by dissolved Fe(II) or from interaction with iron corrosion products such a magnetite or green rusts. The generated hydrogen (up to several MPa) may also inhibit the spent nuclear fuel dissolution at the surface of the fuel via the so called hydrogen effect. In order to probe these effects an autoclave experiment was performed in which a basket with PWR spent nuclear fuel (burnup ∼43 MWd/kgU) was suspended in an autoclave containing a simplified groundwater (10 mM NaCl, 2 mM NaHCO3) with iron powder. The autoclave was sparged and pressurized with argon. Following an initial rise in radionuclide concentrations from dissolution of pre-oxidised phases the U concentration dropped to 3x10-9 M within 76 days, in-line with the solubility of amorphous UO2, expected to form under reducing conditions. Any Cs and Sr release also ceased within 223 days indicating complete transition from dissolution of pre-oxidized phases and instant release fractions to conditions with inhibition of the dissolution of the fuel matrix. Gas phase analysis and pressure monitoring showed a steady build-up of hydrogen at a rate higher than what could be attributed to radiolysis, reaching hydrogen partial pressures of several hundred kPa. The results indicate continuous corrosion of iron, with magnetite as the dominating iron corrosion product.

Convective processes in spent nuclear fuel canisters

2002 ◽  
Author(s):  
Glenn E. McCreery ◽  
Keith G. Condie ◽  
Randy C. Clarksean ◽  
Donald M. McEligot
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Convective Processes ◽  
Nuclear Fuel Canisters

A new approach to spent nuclear fuel recycling for light water reactors in the frame of remix concept

2020 ◽  
Vol 2020 (1) ◽  
pp. 67-77
Author(s):  
Nikita Vladimirivich Kovalyov ◽  
Boris Yakovlevich Zilberman ◽  
Nikolay Dmitrievich Goletskiy ◽  
Andrey Borisovich Sinyukhin
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Light Water Reactors ◽  
New Approach ◽  
Light Water ◽  
Water Reactors
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