Calculation of used nuclear fuel dissolution rates under anticipated Canadian waste vault conditions (original) (raw)
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MRS Proceedings, 1997
ABSTRACTCalculation of used nuclear fuel dissolution rates in a geological disposal vault requires a knowledge of the redox conditions in the vault. For redox conditions less oxidizing than those causing UO2 oxidation to the U3O7 stage, a thermodynamically-based model is appropriate. For more oxidizing redox conditions a kinetic or an electrochemical model is needed to calculate these rates. The redox conditions in a disposal vault will be affected by the radiolysis of groundwater by the ionizing radiation associated with the fuel. Therefore, we have calculated the alpha-, beta- and gamma-dose rates in water in contact with the reference used fuel in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) as a function of cooling time. Also, we have determined dissolution rates of UO2 fuel as a function of alpha and gamma dose rates from our electrochemical measurements. These room-temperature rates are used to calculate the dissolution rates of used fuel at 100°C, the highest t...
Corrosion Science, 2014
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Corrosion of Nuclear Fuel Inside a Failed Waste Container
Encyclopedia of Interfacial Chemistry, 2018
The international concept for the disposal of high level nuclear waste involves multiple barriers including the fuel bundles, durable metal containers, clay buffer and seals, and a deep geologic environment. A key barrier in the Canadian concept is the corrosion resistant container which consists of an outer copper barrier and an inner carbon steel vessel. While designed not to fail, it is judicious to examine the consequences of container failure when the used fuel bundles could be exposed to groundwater leading to their radiolytic corrosion and the release of radionuclides to the groundwater. If failure occurs two corrosion fronts will be established; one on the fuel driven by the alpha radiolysis of water and a second one on the inside of the steel vessel leading to the dissolution of Fe 2+ and the production of H 2. Both 1-dimensional and 2-dimensional models have been developed to determine the influence of redox conditions within the failed container on the fuel corrosion/radionuclide release process. These models take into account water radiolysis, the reaction of radiolytic H 2 O 2 with the fuel both directly and via galvanic coupling to fission product phases in the fuel, the reaction with H 2 (from steel corrosion and water radiolysis) catalyzed on the fuel surface and the scavenging of radiolytic H 2 O 2 by reaction with Fe 2+. The models are described by 1-dimensional diffusion reaction equations and solved numerically using COMSOL Multiphysics, a commercial simulation package based on the finite element model. The 1-dimensional model attempts to calculate corrosion rates on the exposed surface of fuel pellets while the 2-dimensional model attempts to determine the corrosion behaviour within the cracks in the fuel pellet. Despite extensive international efforts the available database is limited. The sensitivity to various reactions in the model will be evaluated and the necessary improvements identified.
Oxidation/Reduction Status of Water Pooled in a Penetrated Nuclear Waste Container - 10265
Nuclear power use is expected to expand in the future and result in hundreds of thousands of metric tons of spent nuclear fuel (SNF). One of the main concerns of nuclear energy use is SNF disposal. Storage in geological repositories is a reasonable solution for the accumulation of SNF. One of the key factors that determine the performance of the proposed geological repository at Yucca Mountain (YM), NV is the release of radionuclides from the engineered barrier system (EBS) by water transport. Over time, EBS, including nuclear waste containers, is expected to fail gradually due to general and localized corrosion. Physical and chemical disturbances in the environment of the repository will lead to different corrosion rates at different locations of the waste packages. Considering the inherent uncertainty of the failure sequence of a waste package, two main failure scenarios are expected: flow through model (penetrations are on the top and bottom of the waste package causing water to flow through it) and bathtub model (penetrations are on the top with the waste package filling with water). In this paper, we consider a bathtub category failed waste container and shed some light on chemical and physical processes that take place in the pooled water and their effects on radionuclide release. We consider two possibilities: temperature stratification of the pooled water versus mixing. Our calculations show that there will be temperature stratification of the pooled water in the lower half of the waste package, and mixing in the upper half. The effect of these situations on oxygen availability and consequently spent fuel alteration and waste container components corrosion is also considered.
Oxidation/Reduction Status of Water Pooled in a Penetrated Nuclear Waste Container
Nuclear power use is expected to expand in the future and result in hundreds of thousands of metric tons of spent nuclear fuel (SNF). One of the main concerns of nuclear energy use is SNF disposal. Storage in geological repositories is a reasonable solution for the accumulation of SNF. One of the key factors that determine the performance of the proposed geological repository at Yucca Mountain (YM), NV is the release of radionuclides from the engineered barrier system (EBS) by water transport. Over time, EBS, including nuclear waste containers, is expected to fail gradually due to general and localized corrosion. Physical and chemical disturbances in the environment of the repository will lead to different corrosion rates at different locations of the waste packages. Considering the inherent uncertainty of the failure sequence of a waste package, two main failure scenarios are expected: flow through model (penetrations are on the top and bottom of the waste package causing water to flow through it) and bathtub model (penetrations are on the top with the waste package filling with water). In this paper, we consider a bathtub category failed waste container and shed some light on chemical and physical processes that take place in the pooled water and their effects on radionuclide release. We consider two possibilities: temperature stratification of the pooled water versus mixing. Our calculations show that there will be temperature stratification of the pooled water in the lower half of the waste package, and mixing in the upper half. The effect of these situations on oxygen availability and consequently spent fuel alteration and waste container components corrosion is also considered.
A multiphase interfacial model for the dissolution of spent nuclear fuel
Journal of Nuclear Materials, 2015
The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO 2 and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO 2 and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO 2 and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H 2 O 2 and O 2). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit fuel degradation to chemical dissolution, which results in radionuclide source term values that are four or five orders of magnitude lower than when oxidative dissolution processes are operative. This paper presents the scientific basis 3 of the model, the approach for modeling used fuel in a disposal system, and preliminary calculations to demonstrate the application and value of the model.
Radiation Induced Spent Nuclear Fuel Dissolution under Deep Repository Conditions
Environmental Science & Technology, 2007
The dynamics of spent nuclear fuel dissolution in groundwater is an important part of the safety assessment of a deep geological repository for high level nuclear waste. In this paper we discuss the most important elementary processes and parameters involved in radiation induced oxidative dissolution of spent nuclear fuel. Based on these processes, we also present a new approach for simulation of spent nuclear fuel dissolution under deep repository conditions. This approach accounts for the effects of fuel age, burn up, noble metal nanoparticle contents, aqueous H 2 and HCO 3concentration, water chemistry, and combinations thereof. The results clearly indicate that solutes consuming H 2 O 2 and combined effects of noble metal nanoparticles and H 2 have significant impact on the rate of spent nuclear fuel dissolution. Using data from the two possible repository sites in Sweden, we have employed the new approach to estimate the maximum rate of spent nuclear fuel dissolution. This estimate indicates that H 2 produced from radiolysis of groundwater alone will be sufficient to inhibit the dissolution completely for spent nuclear fuel older than 100 years.
Spent fuel performance under repository conditions: A model for use in SR-Can
The SKB performance assessment study SR-Can will use a new model for estimating the release rate of radionuclides from the canister after water can enter the canister. This report describes the background to development of the model and the details of the release rate estimates. The fraction of radionuclide inventory that can be released without dissolution of the fuel matrix is called the rapid release fraction (RRF). The RRF for each radionuclide that can be released without matrix dissolution is given as a triangular distribution with a peak at the most likely value. The matrix dissolution rate is estimated from several sets of experimental data. The data show that there is no enhanced dissolution to be expected due to alpha radiolysis under the conditions that will pertain in the SKB repository. The rate of matrix dissolution in the model is given as a triangular distribution with a range from 10 -6 to 10 -8 per year with a peak at 10 -7 per year. King F, Shoesmith D, 2004. Electrochemical studies of the effect of H 2 on UO 2 dissolution, SKB TR-04-20, Svensk Kärnbränslehantering AB. King F, Quinn M J, Miller H H, 1999. The effect of hydrogen and gamma radiation on the oxidation of UO 2 in 0.1 mol⋅dm -3 NaCl solution, SKB TR-99-27, Svensk Kärnbränslehantering AB. Loida A, Grambow B, Geckeis H, 2001. Spent fuel corrosion behavior in salt solution in the presence of hydrogen overpressure, Proc. of ICEM'01: The 8 th Internat. Conf. on Radioactive Waste Management and Environmental Remediation, Bruges, B, September 30-October 4, 2001 (CD-ROM). Neck V, Kim J I, 2001. Solubility and hydrolysis of tetravalent actinides, Radiochim. Acta 89, 1-16. Ollila K, Albinsson Y, Oversby V, Cowper M, 2003. Dissolution rates of unirradiated UO 2 , UO 2 doped with 233 U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method, SKB TR-03-13, Svensk Kärnbränslehantering AB. Ollila K, Oversby V, 2004. Dissolution of of unirradiated UO 2 and UO 2 doped with 233 U under reducing conditions, Posiva/SKB report in preparation. Olander D, 2004. Thermal spike theory of athermal diffusion of fission products due to alpha decay of actinides in spent fuel (UO 2 ), SKB TR-04-17, Svensk Kärnbränslehantering AB. Parks G A, Pohl D C, 1988. Hydrothermal solubility of uraninite, Geochim. Cosmochim. Acta 52, 863-875. Poinssot C, Toulhoat P, Piron J-P, Cappelare C, Desgranges L, Gras J-M, 2000. Operational and scientific questions related to the long term evolution of spent nuclear fuel in dry conditions. Poinssot C, Jegou C, Toulhoat P, Piron J-P, Gras J-M, 2001. A new approach to the RN source term for spent nuclear fuel under geological disposal conditions, Mat. Res Soc. Symp. Proc. Vol 663, 469-476. Poinssot C, Lovera P, Faure M-H, 2002. Assessment of the evolution with time of the instant release fraction of spent nuclear fuel in geological disposal conditions, Mat. Res. Soc. Symp. Proc. Vol 713, 615-623. Rai D, Felmy A R, Ryan J L, 1990. Uranium(IV) hydrolysis constants and solubility product of UO 2 . xH 2 O(am), Inorg. Chem. 29, 260-264. Rai D, Yui M, Moore D A, 2003. Solubility and solubility product at 22°C of UO 2 (c) precipitated from aqueous U(IV) solutions, J. Solution Chem. 32, 1-17. Roudil D, Deschanels X, Trocellier P, Jégou C, Peuget S,Bart J-M, 2004. Helium thermal diffusion in a uranium dioxide matrix, J. Nucl. Mater. 325, 148-158.
Czechoslovak Journal of Physics, 2006
Radiation corrosion in deaerated water/carbon steel systems has been studied. Kinetics of releasing corrosion products into the water and their sorption on the surface of steel tablets is affected by various factors (redox potential, absorbed dose, temperature, irradiation duration). Concentration of corrosion products in the solution was evaluated using various chemical methods. Total concentration of Fe 2+ /Fe 3+ ions in liquid phase was determined by UV/VIS spectrometry. Solid phase was analysed using X-ray diffraction method. Corrosion processes were studied in deaerated distilled water and synthetic granitic water. Corrosion cells consisted of glass ampoules with inserted steel tablets, completely filled with deoxygenated water. Corrosion cells were carefully enclosed so that air diffusion into system during experiment was kept at minimum. 60 Co gamma sources with various dose rates were used for irradiation. The obtained results indicated that radiation noticeably contributed to the formation of corrosion products. Kinetics of radiation corrosion was strongly dependent on the parameters under study. The obtained experimental data should be taken into consideration when predicting effects of corrosion on containers with spent nuclear fuel using mathematical models.
CORROSION, 2003
A mixed-potential model is described to predict the corrosion behavior of used nuclear fuel inside a steel-lined failed Canadian nuclear waste container under anticipated waste vault (repository) conditions. The model accounts for the effects of the alpha radiolysis of water, the precipitation of corrosion products on both the fuel and the carbon steel (CS), and redox reactions between species produced by either radiolysis or corrosion at the fuel surface and by corrosion on the CS liner. The model is based on a series of ten onedimensional reaction-diffusion equations, each describing the mass-transport, precipitation/dissolution, adsorption/ desorption, and redox processes of the ten chemical species included in the model. These equations are solved using finite-difference techniques. A three-layer spatial grid is used, with the two outer layers (of time-varying thickness) representing porous precipitated corrosion products on the uranium dioxide (UO 2) and CS surfaces. The middle layer represents a layer of groundwater solution in the saturated failed containers. Electrochemical rate expressions are used as boundary conditions for species that participate in interfacial electrochemical reactions.