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.

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>

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.

scholarly journals An Integrated Testing Program for a Near-Field Radionuclide Migration Code

1997 ◽  
Vol 506 ◽  
Author(s):  
P. Binks ◽  
A. Fairhurst ◽  
D. Howarth ◽  
P. N. Humphreys ◽  
T. Johnstone ◽  
...  
Keyword(s):  
Near Field ◽  
Field Studies ◽  
Test Cases ◽  
Radionuclide Migration ◽  
Near Surface ◽  
Transport Code ◽  
Liquid Source ◽  
Integrated Testing ◽  
Main Components

ABSTRACTThe DRINK code is a 2D, biogeochemical transport code developed as a research tool to simulate the long term evolution of near surface LLW disposal sites and to generate gaseous and liquid source terms for far field studies. The code was recently upgraded to provide a more generic modelling tool with wider application to radionuclide migration scenarios. During the development of this code, the Generalised Repository Model (GRM), an integrated strategy has been employed to ensure the production of a fully tested, verified and quality assured product. This strategy is based around a code development protocol with three main components: quality assurance and documentation, verification and realism testing. Realism testing includes both peer review and model testing, with the latter including: experimental test cases; natural and anthropogenic analogues; field observations and finally uncertainty and sensitivity analysis. This paper describes the successful application of the protocol to the development and testing of the GRM code with specific emphasis upon verification and realism testing.

Key Phenomena Governing HLW Glass Behavior in the French Deep Geological Disposal

2015 ◽  
Vol 1744 ◽  
pp. 127-138
Author(s):  
Stéphan Schumacher ◽  
Christelle Martin ◽  
Yannick Linard ◽  
Frédéric Angeli ◽  
Delphine Neff ◽  
...  
Keyword(s):  
Sustainable Management ◽  
Near Field ◽  
Steel Corrosion ◽  
Spent Fuel ◽  
Water Leaching ◽  
Residual Rate ◽  
Partially Saturated ◽  
Deep Geological Disposal ◽  
High Level ◽  
Glass Behavior

ABSTRACTAccording to the Planning Act of 28th June 2006, Andra is in charge of ensuring the sustainable management of all radioactive waste generated in France, especially the high-level and long-lived vitrified waste produced from spent fuel recycling.Since 2006, all the studies and research related to the components of HLW cells have been incorporated into a broader R&D program which aims at characterizing and modeling (i) the glass matrix dissolution, (ii) the corrosion of the overpack and the lining, and (iii) the claystone evolution in the near field, considering all the interactions between these surrounding materials. This program, coordinated by Andra, has involved up to eighteen laboratories.After closure of disposal cells and overpack failure, glass alteration is expected to begin in partially saturated conditions due to hydrogen production resulting from carbon steel corrosion in anoxic conditions. Therefore, the glass should at least partially be hydrated by water vapor during thousands of years until complete saturation. A part of the studies aimed to determine the glass behavior in such conditions, the influence of the main parameters (temperature, relative humidity) and consequences of vapor hydration on subsequent radionuclides release by water leaching.In addition, the major part of the work focused on the influence of the environment on glass alteration. The effect of clay pore water on glass alteration rates (initial rate, rate drop and residual rate) was determined and particularly that of pH and magnesium. The nature of steel corrosion products and their interactions with glass alteration were also investigated. All these studies relied on experiments in surface laboratories, in Andra’s underground laboratory, together with natural or archeological analogs and modeling studies.

Reactive transport modelling of a spent fuel repository in a stiff clay formation considering excavation damaged zones

2004 ◽  
Vol 92 (9-11) ◽  
Author(s):  
Laurent De Windt ◽  
Delphine Pellegrini ◽  
Jan van der Lee
Keyword(s):  
Reactive Transport ◽  
Near Field ◽  
Cementitious Materials ◽  
Special Focus ◽  
Spent Fuel ◽  
Transport Modelling ◽  
Pure Diffusion ◽  
Damaged Zone ◽  
Stiff Clay ◽  
Clay Formation

SummaryThe near-field evolution of a spent fuel disposal in a deep stiff clay formation is studied with the coupled chemistry-transport code HYTEC. The study gives an example that such models can be currently used for geometries (2D and 3D) and time scales (100000 y) relevant for performance assessment. The repository consists of short tunnels with MX80 bentonite barriers and cementitious materials for mechanical support. Cesium, iodine and uranium are released from the waste packages considering instant release fractions and congruent dissolution of the fuel pellets. The calculations are carried out with special focus on the excavation damaged zone (EDZ) comparing diffusion process and different advection scenarios in this zone. Cement represents a source of alkaline perturbations but, under the pure diffusion scenario, the alteration of the multi-barrier system remains limited. The presence of the EDZ does not significantly modify radionuclide migration in the pure diffusion case. The advection scenarios, even with very slow flow velocities, illustrate the possibility of preferential pathways through the EDZ for iodine but show almost no effect on the alkaline plume and cesium migration.

Static Dissolution of α-doped UO2 in Boom Clay Conditions: Preliminary Results

2003 ◽  
Vol 807 ◽  
Author(s):  
C. Cachoir ◽  
K. Lemmens
Keyword(s):  
Corrosion Rate ◽  
Performance Assessment ◽  
Dissolution Rate ◽  
Near Field ◽  
Spent Fuel ◽  
Boom Clay ◽  
Preliminary Results ◽  
Icp Ms ◽  
Radiochemical Measurement ◽  
Α Activity

ABSTRACTTo support the performance assessment (PA) calculations for the possible final disposal of spent fuel in the Boom Clay, leach experiments are performed with α-doped UO2 in clay media, simulating various near field ages. The experiments allow to measure the dissolution rate of the UO2 and to determine the assumed relationship between dissolution rate and α-activity. Tests are performed at six different α-activities, simulating various fuel ages, at 25–30°C, for durations ranging from 90 to 720 days, in a glove box with Ar/0.4%CO2 atmosphere. The solutions and solids are analyzed for U isotopes and 238Pu by radiochemical measurement and by ICP-MS. The dissolution rates of the α-doped UO2 are presented for different durations. The resulting corrosion rate is around 300 μgU.m−2.d−1. This is up to 100 times higher than found for similar conditions in the literature. In the presence of clay, there appears to be no correlation between the α-activity and the corrosion rate of the α-doped UO2.

Assessment of Geochemical Containment Properties in the Near-field of a Deep Underground Repository

2002 ◽  
Vol 757 ◽  
Author(s):  
Delphine Pellegrini ◽  
Laurent De Windt
Keyword(s):  
Reactive Transport ◽  
Near Field ◽  
Safety Evaluation ◽  
Distribution Coefficients ◽  
Spent Fuel ◽  
Advective Flow ◽  
Transport Modelling ◽  
Migration Pathway ◽  
Excavation Damaged Zone ◽  
Damaged Zone

ABSTRACTFor safety evaluation of deep repositories, the evolution of chemical containment properties of clayey barriers in spent fuel disposal tunnels are assessed using reactive transport modelling. The disturbances related to cement components are more particularly studied for relevant time scales (100,000 years) and dimensions. Theoretical distribution coefficients (Kd) and maximum concentrations are derived for Cs, Tc and U and their sensitivity to the system evolution estimated. Mineralogical transformations and ion sorption are shown to be interdependent mechanisms controlling the intensity and spatial expansion of the alkaline plume. Simulations for a normal diffusive scenario and an altered one involving an advective flow lead to limited perturbations of the mineralogy and containment properties of the multi-barriers system, but emphasize the possibility of a migration pathway through the excavation damaged zone.

Transient Diffusion from the Near-Field, A 3-D Model

1991 ◽  
Vol 257 ◽  
Author(s):  
A. Bengtsson ◽  
B. Grundfelt ◽  
H. Widén
Keyword(s):  
Source Term ◽  
Near Field ◽  
Three Dimensional ◽  
Bentonite Clay ◽  
Spent Fuel ◽  
Radionuclide Migration ◽  
Transient Transport ◽  
Mass Transport Rate ◽  
D Model ◽  
Low Solubility

ABSTRACTA study of near-field radionuclide migration is presented. The study [1] has been performed in the context of the SKB91 study which is a comprehensive performance assessment of disposal of spent fuel. The objective of the present study has been to enable the assessment of which nuclides can be screened out because they decay to insignificant levels already in the near-field of the repository.A numerical model has been used which describes the transient transport of radionuclides through a small hole in a spent fuel canister imbedded in bentonite clay into a fracture in the rock outside the bentonite. Calculations for more than twenty nuclides, nuclides with both high and low solubility, have been made. The effect of sorption in the bentonite backfill is included. Materials data for bentonite where taken from [2]. The size of the penetration hole was assumed to be constant up to the time when the calculations were terminated, 500 000 years after the deposition. The mass transport rate is controlled by diffusion. The model is three dimensional.This paper describes the geometry of the modelled system, the assumptions concerning the transport resistances at the boundary conditions, the handling of the source term and obtained release curves.

Release of Radionuclides from the Near Field by Various Pathways. The Influence by the Sorption Properties in the Near Field.

1991 ◽  
Vol 257 ◽  
Author(s):  
Leonardo Romero ◽  
Wars Neretnieks ◽  
Luis Moreno
Keyword(s):  
Near Field ◽  
Chemical Properties ◽  
Compartment Model ◽  
Flowing Water ◽  
Spent Fuel ◽  
Fracture Zones ◽  
Difference Model ◽  
3 Dimensional ◽  
Damaged Zone ◽  
Canister Wall

ABSTRACTRadionuclides from a damaged canister for spent fuel will leak out through a damage in the canister wall and spread through the surrounding backfill. They will further migrate into water bearing fractures in the rock, up through the backfill into the damaged zone around the drift and into the drift itself. Some substance may also diffuse through the rock to adjacent fracture zones. Underway the nuclides will sorb on the materials along the transport paths. This very complex and variable transport geometry has been modelled using a compartment model which is based on simplifying a full 3 dimensional integrated finite difference model. The simplifications are supplemented by introducing analytical and semianalytical solutions at sensitive locations such as entrances and exits from damages and fractures and in the flowing water. The model has been tested against full 3D solutions with good results. Sample calculations are presented and discussed for a nuclide with the chemical properties of Pu-239.

Near-Field Modelling in the Safety Assessment SR-Can

2008 ◽  
Vol 1107 ◽  
Author(s):  
C. Fredrik Vahlund
Keyword(s):  
Numerical Modelling ◽  
Input Data ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Fractured Rock ◽  
Near Field ◽  
Spent Fuel ◽  
Radionuclide Migration ◽  
Field Modelling ◽  
Swelling Clays

AbstractSpent nuclear fuel from the Swedish energy programme will be stored in an underground repository situated in saturated fractured rock at a depth of approximately 500 m. This paper describes numerical simulations of radionuclide migration in the near-field (consisting of a canister filled with spent fuel and an engineered system backfilled with swelling clays) for the recently completed safety assessment SR-Can [1] using a Matlab / Simulink code. Handling of input data for the models from the site descriptive programme from on-going investigations at two candidate sites and the numerical modelling concept are discussed.

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