Near Field Performance of the Advanced Cold Process Canister

1990 ◽  
Vol 212 ◽  
Author(s):  
Lars O. Werme ◽  
Jukka-Pekka Salo
Keyword(s):  
Performance Evaluation ◽  
Carbon Steel ◽  
Service Life ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Field Performance ◽  
Load Bearing ◽  
Bearing Element

ABSTRACTA near-field performance evaluation of an advanced cold process canister for spent nuclear fuel has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a carbon steel canister as a load bearing element, with an outer corrosion shield of copper. The canister design was originally designed by TVO. In the analysis, internal (ie corrosion processes from the inside of the canister) as well as external processes (mechanical and chemical) have been considered both prior to and after canister breach. Throughout the analysis, present day underground conditions have been assumed to persist during the service life of the canister.

Radionuclide behaviour in the near-field of a geological repository for spent nuclear fuel

2012 ◽  
Vol 100 (8-9) ◽  
pp. 699-713 ◽  
Author(s):  
Volker Metz ◽  
Horst Geckeis ◽  
Ernesto Gonzáles-Robles ◽  
Andreas Loida ◽  
Christiane Bube ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Geological Repository

Fracture smectite as a long-term sink for natural radionuclides – indications from unusual U-series disequilibria

2000 ◽  
Vol 88 (9-11) ◽  
Author(s):  
N. Marcos ◽  
J. Suksi ◽  
H. Ervanne ◽  
K. Rasilainen
Keyword(s):  
Nuclear Fuel ◽  
Chemical Stability ◽  
Spent Nuclear Fuel ◽  
Activity Ratio ◽  
Near Field ◽  
Natural Radionuclides ◽  
Mineral Component ◽  
Considerable Part ◽  
Preferential Loss

Occurrence of natural U in fracture smectite (main mineral component of bentonite) was studied as an analogue to radionuclide behaviour in the near-field of spent nuclear fuel repository. Elevated U content (57 ppm) was observed in fracture smectite sampled from the surface of water-carrying fracture in granite pegmatite at a depth of 70 m. The current groundwater conditions are oxidising at the sampled point. The U-234/U-238 activity ratio (AR) measured in the bulk U and in its sequentially extracted phases, displays unusually low value (around 0.30). Low AR indicates preferential loss of the U-234 isotope from the system. Because the U-234 loss can also be seen in the Th-230/U-234 activity ratio (clearly over 1), the selective removal of the U-234 isotope must have taken place more recently than what is needed to equilibrate Th-230/U-234 pair (i.e. 350000 a). To explain the selective U-234 loss from the smectite we postulate that bulk U is in reduced +4 form and a considerable part of the U-234 isotope in easily leachable oxidised +6 form. This study suggests that the long-term chemical stability of the bulk U in the smectite is due to irreversible fixation of U in the reduced +4 form.

Preliminary Calculations of Radionuclide Release From Near-Field to Biosphere Using Computer Code Goldsim

2003 ◽  
Author(s):  
S. Prva´kova´ ◽  
Juraj Dˇu´ran
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Slovak Republic ◽  
Near Field ◽  
Compartment Model ◽  
Computer Code ◽  
Geological Disposal ◽  
Rock Formations ◽  
Groundwater Supply ◽  
Radionuclide Release

This paper presents preliminary calculations for the performance assessment of the hypothetical disposal facility for spent nuclear fuel in the Slovak Republic. In the concept of geological disposal, two host rock formations are under consideration: granite and clay. Slovakia is now in site selection phase therefore the lack of input data for calculations is compensated from the international reports [1, 2]. The calculations of radionuclide release were made for the spent nuclear fuel disposed in a hypothetical underground repository situated in clay host formation. One-dimensional mathematical model of the radionuclide release through the engineered and natural barrier system up to one million years was simulated using a flexible compartment model tool Goldsim. Both deterministic and probabilistic calculations have been applied. The input parameters for probabilistic calculations are selected from the probability density functions (PDF) using Monte Carlo method. Radionuclide release to the biosphere is considered via contaminated groundwater supply taken from a deep well. Three different exposure pathways were considered: consumption of the root vegetables, consumption of meat and drinking of water from well. The results of preliminary assessment of radiological consequences are expressed in terms of release rates on the aquifer-well interface together with the corresponding individual annual doses for average member of the critical group of adults living in the vicinity of the repository. The regulatory limits regarding the annual doses for the concept of geological disposal in Slovakia are under preparation.

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>

Integrated Model of Korean Spent Fuel and High Level Waste Disposal Options

2009 ◽  
Author(s):  
Yongsoo Hwang ◽  
Ian Miller
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Near Field ◽  
Integrated Model ◽  
High Level Waste ◽  
Spent Fuel ◽  
Parameter Uncertainties ◽  
Design Concepts ◽  
High Level

This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21’st century. The model addresses alternative design concepts for disposal of SNF of different types (CANDU, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model’s results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses.

Near-field behavior of 99Tc during the oxidative alteration of spent nuclear fuel

2000 ◽  
Vol 278 (2-3) ◽  
pp. 225-232 ◽  
Author(s):  
Fanrong Chen ◽  
Peter C. Burns ◽  
Rodney C. Ewing
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Field Behavior

Corrosion behavior of spent nuclear fuel in high pH solutions – Effect of hydrogen

2012 ◽  
Vol 1475 ◽  
Author(s):  
A. Loida ◽  
R. Gens ◽  
C. Bube ◽  
K. Lemmens ◽  
C. Cachoir ◽  
...  
Keyword(s):  
Detection Limit ◽  
Carbon Steel ◽  
Nuclear Fuel ◽  
Corrosion Behavior ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Detection Limits ◽  
Type Solution ◽  
Corrosion Studies ◽  
Fractional Release

ABSTRACTIn accordance with the Belgian “supercontainer design”, spent nuclear fuel (SNF) will be encapsulated in carbon steel canisters, surrounded by a concrete overpack for disposal in poorly-indurated clay. After re-saturation of the barriers by porewater, interactions with the concrete will result in solutions rich in NaOH, KOH and Ca(OH)2. Corrosion studies of SNF in ECW-type solution (Evolved Cement Water) and YCWCa-type solution (Young Cement Water with Ca) were performed under externally applied H2 overpressures over 426 days. Directly after H2 application, Tc concentrations decreased from >10-8 M to concentrations below detection limit. Based on the fractional release of selected fission products, low matrix dissolution rates of ~10-8/day were found in both experiments. U concentrations decreased finally to 1.5•10-9 M (YCWCa) and to 2.1•10-10 M (ECW), respectively. Am, Np and Pu concentrations were found throughout the experiments below their detection limits indicating an effective retention process.

Assessment of the relevance of Coffinite formation within the near-field environment of spent nuclear fuel geological disposals

2006 ◽  
Vol 932 ◽  
Author(s):  
V. Robit-Pointeau ◽  
C. Poinssot ◽  
P. Vitorge ◽  
B. Grambow ◽  
D. Cui ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Thermodynamic Data ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
New Mineral ◽  
Short Term ◽  
Geological Disposal ◽  
Field Environment ◽  
Permanent Presence

ABSTRACTExperiments were performed in anoxic gloves box in an attempt to synthesise Coffinite both in representative near-field conditions, and in conditions which were expected to favour its precipitation according to thermodynamic calculations. The experimental results did not confirm the predictions. However, a new mineral was observed instead of Coffinite. In addition, accurate characterization of various natural samples demonstrate the permanent presence of U(VI) within Coffinite contradictory to its theoretical composition. Our observations raise the question on the validity and applicability of available –actually estimated- thermodynamic data of Coffinite. Based on kinetic hindrance of Coffinite formation, coffinitization of spent nuclear fuel in geological disposal is not anticipated to be a dominant short term process.

scholarly journals A Science-Based Approach to Understanding Waste Form Durability in Open and Closed Nuclear Fuel Cycles

2006 ◽  
Vol 985 ◽  
Author(s):  
M. T. Peters ◽  
R. C. Ewing
Keyword(s):  
Nuclear Fuel ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Near Field ◽  
The United States ◽  
Waste Form ◽  
Scale Model ◽  
United States Department ◽  
Waste Forms

ABSTRACTThere are two compelling reasons for understanding source term and near-field processes in a radioactive waste geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are degraded, the waste form is a primary control on the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years. The United States Department of Energy's Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: a) SNF dissolution mechanisms and rates; b) formation and properties of U6+-secondary phases; c) waste form–waste package interactions in the near-field; and d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of source term and near-field processes. This integrated model will be used to provide a basis for understanding the behavior of the source term over long time periods (greater than 105 years). Such a fundamental and integrated experimental and modeling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms “tailored” to specific geologic settings.

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