Modeling spent nuclear fuel alteration and radionuclide migration in disposal conditions

2006 ◽  
Vol 94 (9-11) ◽  
Author(s):  
Laurent de Windt ◽  
H. Schneider ◽  
C. Ferry ◽  
H. Catalette ◽  
V. Lagneau ◽  
...  
Keyword(s):  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Transport Model ◽  
Near Field ◽  
Radioactive Decay ◽  
Steel Corrosion ◽  
Spent Fuel ◽  
Fuel Pellets ◽  
Physico Chemical ◽  
Integrated Performance

A physico-chemical model developed for spent fuel alteration was integrated in a global reactive transport model of a spent fuel disposal system, considering both decaying and stable isotopes, corroded steel canisters, bentonite backfills and a clayey host-rock. Fuel evolution took into account radiolytic-enhanced corrosion and long-term solubility-controlled dissolution as well as instantaneous release fractions. The calculations show that spent-fuel dissolution has no significant alteration effect on the near-field components except an oxidizing plume in the vicinity of the waste packages. The dissolved uranyl species, partly precipitate as schoepite on the fuel pellets, and partly diffuse in the near-field where magnetite and pyrite reduce U(VI) to yield uraninite precipitation. Under disposal conditions, preliminary calculations indicate that steel corrosion may generate sufficient dissolved hydrogen as to react with radiolytic oxidants and inhibit fuel dissolution. The formation of a protective schoepite layer could also reduce the alteration of fuel pellets. Radionuclides migration (Am, Cs, I) in the near-field is discussed in a second stage discriminating between sorption, precipitation and radioactive decay processes. The migration of Cs is translated in terms of cumulative activity profiles useful for integrated performance assessment.

The Influence of Ligands and Precipitates on the Release of Nuclides from the Near Field under Natural Repository Conditions

2000 ◽  
Vol 663 ◽  
Author(s):  
L. Liu ◽  
I. Neretnieks
Keyword(s):  
Uranium Oxide ◽  
Spent Nuclear Fuel ◽  
Transport Model ◽  
Radiation Power ◽  
Near Field ◽  
Reference Scenario ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Fuel Surface ◽  
Fuel Pellets

ABSTRACTOnce groundwater intrudes into a damaged canister and wets the spent fuel pellets, radiation emitted from the spent nuclear fuel splits nearby water into oxidizing and reducing species. This may lead to an oxidizing condition near the fuel pellets. As a result, uranium oxide that makes up the fuel matrix will become more soluble, and the incorporated radionuclides will be released more rapidly. The dissolution process is, however, a dynamic one that can be influenced by many factors. Of great importance are the radiation power of the fuel matrix, the concentration of ligands near the fuel surface, and the transport resistance of the near field. Consequently, the escape of nuclides from the damaged canister is dominated mainly by the intrusion of ligands, and the precipitation/dissolution of secondary phases within the fuel rods. To investigate the possible effects of ligands and precipitates, a coupled dissolution and transport model, which includes the barrier effect of the Zircaloy claddings, is developed. The application of the model to a SKB-specified reference scenario indicates that by far the largest fraction of the oxidized uranium will reprecipitate within the canister. This may significantly decrease the fuel surface available for oxidation and the water available for radiolysis. Subsequently, much less fuel matrix will be dissolved and much less of the other nuclides will be released. Simulations further identify that carbonate and silicate have the greatest influences on the formation of secondary phases, and on the release of nuclides, under natural repository conditions.

Reactive Transport Modeling of Radionuclide Source-Terms in an Underground Spent Fuel Disposal

2003 ◽  
Vol 807 ◽  
Author(s):  
L. De Windt ◽  
H. Catalette ◽  
J. M. Gras
Keyword(s):  
Reactive Transport ◽  
Transport Model ◽  
Near Field ◽  
Disposal Site ◽  
Spent Fuel ◽  
Fuel Pellets ◽  
Backfill Materials ◽  
Global Equilibrium ◽  
Stiff Clay ◽  
Failure Scenario

ABSTRACTThe reactive transport model HYTEC was used to simulate the migration over 100,000 years of cesium, americium and uranium released from spent fuel packages in the near-field components of an underground stiff clay disposal site. A global equilibrium thermodynamic approach including kinetic control of the spent fuel pellets was used with instantaneous release fractions and congruent dissolutions of the rim and the core zones. A failure scenario of the waste package after 10,000 years was considered with magnetite as the main corrosion product. The retention properties of magnetite and the different effects of bentonite and cementitious backfill materials were specifically analysed.

Near field evolution of a spent fuel repository in an argillaceous rock formation and impact on radionuclide migration

2020 ◽  
Author(s):  
Vanessa Montoya ◽  
Orlando Silva ◽  
Emilie Coene ◽  
Jorge Molinero ◽  
Renchao Lu ◽  
...  
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Spent Fuel ◽  
Exchange Reactions ◽  
Radionuclide Migration ◽  
High Level

<p>In August 2015, the German government approved the national programme for the responsible and safe management of spent nuclear fuel (SNF) and radioactive waste proposed by the Federal Ministry for the Environment, Nature Conservation, Building and Reactor Safety (BMU). The assumption is that about ~ 1 100 storage casks (10 500 tons of heavy metal) in the form of spent fuel assemblies will be generated in nuclear power plants and will have to be disposed. However, a decision on the disposal concept for high-level waste is pending and an appropriate solution has to be developed with a balance in multiple aspects. All potential types of host rocks, clay and salt stones as well as crystalline formations are under consideration. In the decision process, evaluation of the risk of different waste management options and scenarios play an enormous role in the discussion. Coupled physical and chemical processes taking place within the engineered barrier system of a repository for high-level radioactive waste will define the radionuclide mobility/retention and the possible radiological impact. The objective of this work is to assess coupled processes occurring in the near-field of a generic repository for spent nuclear fuel in a high saline clay host rock, integrating complex geochemical processes at centimetre-scale. The scenario considers that radionuclides can be released during a period of thousands of years after full saturation of the bentonite barrier and the thermal phase.</p><p>Transport parameters and the discretization of the system, are implemented in a 2D axisymmetric geometry. The multi-barrier system is emplaced in clay and a solubility limited source term for the selected radionuclides is assumed. Kinetics and chemical equilibria reactions are simulated using parameters obtained from experiments. Additionally, porosity changes due to mineral precipitation/dissolution and feedback on the effective diffusion coefficient are taken into account. Protonation/deprotonation, ion exchange reactions and radionuclide inner-sphere sorption is considered.</p><p>Numerical simulations show, that, when the canister corrosion starts, the redox potential decreases, magnetite precipitates and H<sub>2</sub> is formed. Furthermore, the aqueous concentration of Fe(II) increases due to the presence of magnetite. By considering binding to montmorillonite via ion exchange reactions, the bentonite acts as a sink for Fe(II). Additionally, magnetite forms a chemical barrier offering significant sorption capacity for many radionuclides. Finally, a decrease of porosity in the bentonite/canister interface leads to a further deceleration of radionuclide migration. Due to the complexity of reactive transport processes in saline environments, benchmarking of reactive transport models (RTM) is important also to build confidence in those modelling approaches. Development of RTM benchmark procedures is part of the iCROSS project (Integrity of nuclear waste repository systems - Cross-scale system understanding and analysis) funded by both the Helmholtz Association and the Federal Ministry of Education and Research (BMBF).</p><p> </p>

Modelling Radionuclide Mobility of Spent Fuel in a Deep Geological Repository Located in a Clay Bedrock

2009 ◽  
Author(s):  
Lara Duro ◽  
Abel Tamayo ◽  
Jordi Bruno ◽  
Aurora Marti´nez-Esparza
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Radioactive Decay ◽  
Secondary Phase ◽  
Diffusive Transport ◽  
Deep Geological Repository ◽  
Spent Fuel ◽  
Geological Repository ◽  
Source Term Model

Source term models are widely used to assess the behaviour of spent nuclear fuel after final disposal. However, most models do not take into account some phenomena which are expected to control the transport of radionuclides through the near field. Some uncertainties arise from this fact, thus making it difficult to obtain proper simulations of radionuclide behaviour in the near field. In this work, we have used a compartmental code to build up an integrated source term model in an attempt to overcome the abovementioned drawbacks. The model developed takes into account radiolytically-mediated matrix dissolution, radioactive decay chains, diffusive transport, and retardation by sorption and secondary phase precipitation, among other processes. In addition, this model has been used to estimate radionuclide mobility from spent fuel located in a conceptual clay geological repository.

scholarly journals Reactive transport calculations to evaluate sulphide fluxes in the near-field of a SF/HLW repository

2019 ◽  
Vol 98 ◽  
pp. 10005
Author(s):  
Marek Pękala ◽  
Paul Wersin ◽  
Veerle Cloet ◽  
Nikitas Diomidis
Keyword(s):  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Waste Containment ◽  
Geological Repository ◽  
High Level ◽  
Reducing Bacteria ◽  
Hlw Repository

Radioactive waste is planned to be disposed in a deep geological repository in the Opalinus Clay (OPA) rock formation in Switzerland. Cu coating of the steel disposal canister is considered as potential a measure to ensure complete waste containment of spent nuclear fuel (SF) and vitrified high-level waste (HLW) or a period of 100,000 years. Sulphide is a potential corroding agent to Cu under reducing redox conditions. Background dissolved sulphide concentrations in pristine OPA are low, likely controlled by equilibrium with pyrite. At such concentrations, sulphide-assisted corrosion of Cu would be negligible. However, the possibility exists that sulphate reducing bacteria (SRB) might thrive at discrete locations of the repository’s near-field. The activity of SRB might then lead to significantly higher dissolved sulphide concentrations. The objective of this work is to employ reactive transport calculations to evaluate sulphide fluxes in the near-field of the SF/HLW repository in the OPA. Cu canister corrosion due to sulphide fluxes is also simplistically evaluated.

Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

2008 ◽  
Vol 1104 ◽  
Author(s):  
Claude Degueldre ◽  
Wolfgang Wiesenack
Keyword(s):  
Near Field ◽  
Fission Products ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Fuel Pellets ◽  
Reducing Conditions ◽  
Deep Geological Disposal

AbstractA plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.

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.

Spent fuel alteration model integrating processes of different time-scales

MRS Advances ◽  
2020 ◽  
Vol 5 (3-4) ◽  
pp. 159-166
Author(s):  
O. Riba ◽  
E. Coene ◽  
O. Silva ◽  
L. Duro
Keyword(s):  
Reactive Transport ◽  
Transport Model ◽  
Beta Radiation ◽  
Comsol Multiphysics ◽  
Water Radiolysis ◽  
Spent Fuel ◽  
Subsequent Reduction ◽  
Reactive Transport Model ◽  
Safety Assessments ◽  
Different Time Scales

ABSTRACTA 1D reactive transport model has been implemented in iCP (interface COMSOL Multiphysics and PhreeqC) to assess the corrosion of Spent Fuel (SF), considered as homogeneous UO2(am,hyd) doped with Pd. The model couples: i) generation of water radiolysis species by alpha and beta radiation considering the complete water radiolysis system with the kinetic reactions involving: H+, OH-, O2, H2O2, H2, HO2-, HO2·, O·, O-, O2-, H·, ·OH and e- ii) processes occurring in the spent fuel surface: oxidative dissolution reactions of UO2(am,hyd) and subsequent reduction of oxidized fuel, considering H2 activation by Pd, and iii) corrosion of Fe(s) in oxic and anoxic conditions. Process i) has been implemented in COMSOL and processes ii) and iii) have been implemented in PHREEQC with their kinetic constants being calibrated with different sets of experimental data published in the open literature. The model yields a UO2(am,hyd) dissolution rates similar to the values selected in safety assessments.

Nuclear Waste

2014 ◽  
Author(s):  
Michael H. Fox
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Spent Nuclear Fuel ◽  
Nuclear Fission ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Fuel Pellets ◽  
Water Filling ◽  
Nuclear Fuel Waste ◽  
Ranch House

I gazed over the railing into the crystal clear cooling pool glowing with blue Cherenkov light caused by particulate radiation traveling faster than the speed of light in water. I can see a matrix of square objects through the water, filling more than half of the pool. It looks like you could take a quick dip into the water, like an indoor swimming pool, but that would not be a good idea! It is amazing to think that this pool, about the size of a ranch house, is holding all of the spent fuel from powering the Wolf Creek nuclear reactor in Burlington, Kansas, for 27 years. The reactor was just refueled about a month before my visit, so 80 of the used fuel rod assemblies were removed from the reactor and replaced with new ones. The used fuel rods were moved underwater into the cooling pool, joining the approximately 1,500 already there. There is sufficient space for the next 15 years of reactor operation. There is no danger from standing at the edge of this pool looking in, though the levels of radon tend to be somewhat elevated and may electrostatically attach to my hard hat, as indeed some did. What I am gazing at is what has stirred much of the controversy over nuclear power and is what must ultimately be dealt with if nuclear power is to grow in the future—the spent nuclear fuel waste associated with nuclear power. What is the hidden danger that I am staring at? Am I looking at the unleashed power of Hephaestus, the mythical Greek god of fi re and metallurgy? Or is this a more benign product of energy production that can be managed safely? What exactly is in this waste? And is it really waste, or is it a resource? To answer that question, we have to understand the fuel that reactors burn. The fuel rods that provide the heat from nuclear fission in a nuclear reactor contain fuel pellets of uranium, an element that has an atomic number of 92 (the number of protons and also the number of electrons).

The Near-Field Transport Code Tullgarn and its Use in Performance Assessment

1992 ◽  
Vol 294 ◽  
Author(s):  
Patrik Sellin ◽  
Nils Kjellbert
Keyword(s):  
Performance Assessment ◽  
Near Field ◽  
Steel Corrosion ◽  
List Type ◽  
Spent Fuel ◽  
Radionuclide Migration ◽  
Transport Code ◽  
Damaged Zone ◽  
Chain Decay ◽  
Resistance Network Model

ABSTRACTThe near-field radionuclide migration code Tullgarn has been developed for performance assessment purposes. As a part of the PROPER-code package it has been successfully applied in the SKB 91 safety analysis.The features and processes included in the code are:- Radioactive chain decay- Different canister failure mechanisms (copper corrosion from sulphide attack, steel corrosion, internal overpressure and initially defective canisters) - Spent fuel dissolution. The model is based on the assumption that the dissolution rate is proportional to the α-dose rate- Transport calculations are done with a resistance-network model. Tullgarn calculates the stationary release of radionuclides from a defect in the canister through the buffer and out into a fracture in the rock or up to the damaged zone under the deposition tunnel.Tullgarn can be used as a stand-alone model for near-field release calculations or as a submodel in an integrated assessment. In the SKB 91 analysis, Tullgarn gave the source term to the far-field model.

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