Oxidation/Reduction Status of Water Pooled in a Penetrated Nuclear Waste Container (original) (raw)
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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.
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.
CORROSION, 2018
Models for the corrosion of spent nuclear fuel (fission and actinide-doped uranium dioxide) provide the essential source term for the release of radionuclides from within a failed nuclear waste container in a deep geologic repository. Although redox conditions within a repository are expected to be anoxic, exposure of the fuel to groundwater will cause the generation of oxidants at the fuel surface, leading to its corrosion and the release of radionuclides. The influence of these oxidants will be partially mitigated by the anoxic corrosion of the inner walls of the steel container to produce the oxidant scavengers, Fe 2+ and H 2. This review summarizes the development of a finite element model developed to determine the influence of the various redox-controlling species (H 2 O 2 , Fe 2+ , H 2). Both one-dimensional and two-dimensional models are described, with the latter required to take into account the fractured geometry of the fuel.
Limitations on Radionuclide Release From Partially Failed Containers
Materials Research Society Symposium Proceedings, 2008
In the long run, nuclear waste packages at the Yucca Mountain repository are likely to evolve into a combination of corroded materials mixed with relicts of intact Alloy-22 and other waste package materials. Different rates of corrosion, due to physical and chemical disturbances in the environment of the repository, will lead to different times of penetration between waste packages and at different locations on the same waste package. Radionuclides are released from waste packages by dissolution and transport in water. In this paper, we shed some light on the effect of residual heat release, and other physical processes that take place in the waste package during penetration times, on radionuclide release. We develop a flow-through conceptual model for a probable serious failure in which multiple penetrations allow water to flow through a partially failed waste container. This model demonstrates that evaporation at hotter protected areas creates a capillary pressure gradient that causes water to flow with its dissolved and suspended contents toward these relict protected areas, effectively preventing radionuclide release. We derive a dimensionless group to estimates the minimum size of the covered areas required to sequester radionuclides and prevent release, and explore the implication of the flow-through model on the Yucca Mountain repository performance.
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.
Corrosion Science, 2014
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Calculation of used nuclear fuel dissolution rates under anticipated Canadian waste vault conditions
Journal of Nuclear Materials, 1997
Dissolution rates of UO fuel are determined as a function of alpha and gamma dose rates. These room-temperature rates 2 are used to calculate the dissolution rates of used fuel at 1008C. Also, the alpha, beta and gamma dose rates in water in contact with the reference used fuel are calculated as a function of cooling time. These results are used to calculate used CANDU fuel dissolution rates as a function of time since emplacement in a defective copper container for the Canadian Nuclear Fuel Waste Management Program. It is shown that beta radiolysis of water is the main cause of oxidation of used CANDU fuel in a failed container and that the use of a corrosion model is required for ; 1000 a of emplacement in the waste vault. The results obtained here can be adopted to calculate used nuclear fuel dissolution rates for other waste management programs. q 1997 Elsevier Science B.V. h 4 93 7 w x 17 . However, any model of fuel dissolution within a 0022-3115r97r$17.00 q 1997 Elsevier Science B.V. All rights reserved. Ž . PII S 0 0 2 2 -3 1 1 5 9 7 0 0 2 7 2 -9
JOM, 2014
Metallic engineered barriers must provide a period of absolute containment to high-level radioactive waste in geological repositories. Candidate materials include copper alloys, carbon steels, stainless steels, nickel alloys, and titanium alloys. The national programs of nuclear waste management have to identify and assess the anticipated degradation modes of the selected materials in the corresponding repository environment, which evolves in time. Commonly assessed degradation modes include general corrosion, localized corrosion, stress-corrosion cracking, hydrogen-assisted cracking, and microbiologically influenced corrosion. Laboratory testing and modeling in metallurgical and environmental conditions of similar and higher aggressiveness than those expected in service conditions are used to evaluate the corrosion resistance of the materials. This review focuses on the anticipated degradation modes of the selected or reference materials as corrosion-resistant barriers in nuclear repositories. These degradation modes depend not only on the selected alloy but also on the near-field environment. The evolution of the near-field environment varies for saturated and unsaturated repositories considering backfilled and unbackfilled conditions. In saturated repositories, localized corrosion and stress-corrosion cracking may occur in the initial aerobic stage, while general corrosion and hydrogen-assisted cracking are the main degradation modes in the anaerobic stage. Unsaturated repositories would provide an oxidizing environment during the entire repository lifetime. Microbiologically influenced corrosion may be avoided or minimized by selecting an appropriate backfill material. Radiation effects are negligible provided that a thick-walled container or an inner shielding container is used.