A Note on Radionuclide Transport by Gas Bubbles

1996 ◽  
Vol 465 ◽  
Author(s):  
Ivars Neretnieks ◽  
Märta-Lena Ernstson
Keyword(s):  
Carrying Capacity ◽  
Colloidal Particles ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Ground Surface ◽  
Gas Bubbles ◽  
Escape Rate ◽  
Fuel Gas ◽  
Buffer Material ◽  
Bubble Transport

ABSTRACTIn a repository for spent nuclear fuel, gas generated by corrosion of the iron in the canister may form small bubbles that will escape and rise to the ground surface. Colloidal particles may attach to the surface of the bubbles and be carried by them. If the colloids are supplied by the montmorillonite clay of the buffer material surrounding the canister, the clay can be carried away. Nuclides sorbed in the clay can be carried with the bubbles. We have estimated the carrying capacity of the gas of the clay particles and the escape rate of nuclides carried by the gas bubbles. The latter is also compared to the escape rate by the conventional escape mechanisms from the near field. We have further estimated the detachment of the nuclides from the clay and their sorption onto the fracture surfaces of the rock as well as their uptake by diffusion into the rock matrix along the bubble transport paths.The present paper is speculative and uses some hypothetical assumptions. Although the processes that are modelled are known to exist there is not enough known of several of them to quantify them accurately. The carrying capacity of the gas used in the calculations is an upper bound and probably very much exaggerated. Even so, the consequences are minor for the release of radionuclides.

Some Important Mechanisms and Processes in the Near Field of the Swedish Repository for Spent Nuclear Fuel

1992 ◽  
Vol 294 ◽  
Author(s):  
Ivars Neretnieks
Keyword(s):  
Molecular Diffusion ◽  
Spent Nuclear Fuel ◽  
Dominant Role ◽  
Near Field ◽  
Water Flow Rate ◽  
Crystalline Rock ◽  
Diffusive Transport ◽  
Actual Design ◽  
Natural Analogues ◽  
And Performance

ABSTRACTIn repositories for nuclear waste there are many processes that will be instrumental in damaging the canisters and releasing the nuclides. Based on experiences from studies of the performance of repositories and of an actual design, the major mechanisms influencing the integrity and performance of a repository are described and discussed. The paper addresses only conditions in crystalline rock repositories. The low water flow rate in fractures and channels plays a dominant role in limiting the interaction between water and waste. Molecular diffusion in the backfill and rock matrix, as well as in the mobile water, is an important transport process, but actually limits the exchange rate because diffusive transport is slow. Solubility limits of both waste matrix and of individual nuclides are also important. Complicating processes include alpha-radiolysis, which may change the water chemistry in the near-field. The sizes and locations of water flowpaths and damages in the canisters considerably influence the release rates. Uncertainties in data are large. Nevertheless the system is very robust in the sense that practically no reasonably conceivable assumptions or data will lead to large nuclide releases. Several natural analogues have been found to exhibit similarities with a waste repository and help to validate concepts and to increase our confidence that all major issues have been considered.

Predictions of residual carrying-capacities for fire and near-field blast-damaged reactive powder concrete-filled steel tube columns

2018 ◽  
Vol 9 (4) ◽  
pp. 525-553 ◽  
Author(s):  
Wanxiang Chen ◽  
Zixin Zhou ◽  
Huihui Zou ◽  
Zhikun Guo
Keyword(s):  
Carrying Capacity ◽  
Large Scale ◽  
Blast Resistance ◽  
Near Field ◽  
Theoretical Method ◽  
Steel Tube ◽  
Reactive Powder Concrete ◽  
Concrete Filled Steel Tube ◽  
Damage Indices ◽  
Reactive Powder

An approximate approach is developed to estimate the residual carrying-capacities of fire and near-field blast-damaged reactive powder concrete-filled steel tube columns. The single-degree-of-freedom model is employed to calculate the initial deflections of fire-damaged reactive powder concrete-filled steel tube columns subjected to axial and blast-induced transverse loads, and then a modified formula including double coefficient is further proposed to predict the ultimate resistance. Then, a series of blast-resistance and load carrying-capacity tests on six large-scale reactive powder concrete-filled steel tube columns are conducted to validate the suitability of theoretical method presented in this article. Blast tests demonstrate that the blast-resistances of reactive powder concrete-filled steel tube columns are more sensitive to fire durations than to scale distances. In addition, it is indicated that ISO-834 standard fire exposures cause significant degradations of material properties and have remarkable effects on the residual carrying-capacities of reactive powder concrete-filled steel tube columns. No local bucking and burst could be observed in the residual carrying-capacity tests; also, there are no visible hinge-like deformations in the mid-span area, and the excellent fire-resistances and blast-resistances of reactive powder concrete-filled steel tube columns are experimentally verified. Analytical results show that the predicted axial load capacities of six reactive powder concrete-filled steel tube columns are in good agreement with experimental data. All damage indices of the test specimens are within 0.8, meaning only minor to severe damage is done to the reactive powder concrete-filled steel tube column during fire and blast attacks, which is consistent with the test results.

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.

Coupling Swelling and Water Retention Processes in Compacted Bentonite

2011 ◽  
Vol 6 (1) ◽  
Author(s):  
Antti Lempinen
Keyword(s):  
Water Retention ◽  
Spent Nuclear Fuel ◽  
Effective Strain ◽  
Swelling Pressure ◽  
Model Parameters ◽  
Confined Water ◽  
Swelling Properties ◽  
Compacted Bentonite ◽  
Buffer Material ◽  
Swelling Factor

Compacted bentonite is the main candidate for buffer material in several plans for spent nuclear fuel repositories. One of its important properties is high swelling capacity, which is caused by interaction between water molecules and exchangeable cations. This interaction makes bentonite behave differently from capillary materials. In this article, a model for thermo-hydro-mechanical state of partially water saturated bentonite is presented. It couples the water retention and swelling properties with introduction of the swelling factor in effective strain. The Helmholz energy density determines the state with a relatively small set of independent parameters: swelling pressure, swelling factor, maximum confined water content and the reference state. The model parameters are determined from experimental data for FEBEX bentonite, and as a simple consistency check, confined suction curves are calculated and compared to test results. Consistency of the model with observations on nano- and microscale of bentonite is also discussed.

New insight studies of the secondary phase formation under repository conditions

MRS Advances ◽  
2020 ◽  
Vol 5 (11) ◽  
pp. 539-547
Author(s):  
N. Rodríguez-Villagra ◽  
L.J. Bonales ◽  
J. Cobos
Keyword(s):  
Raman Spectroscopy ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Aqueous Media ◽  
Secondary Phase ◽  
Ambient Atmosphere ◽  
Secondary Phases ◽  
Γ Radiation ◽  
Radiation Fields ◽  
Groundwater Environment

ABSTRACTIn a deep geological repository (DGR) scenario, uranium oxidized in aqueous systems will be stabilized as UO22+ (hexavalent uranium), as a consequence of tetravalent uranium oxidation by radiolytic byproducts. Uranyl cationic species (UO22+) in different speciation forms are expected to be found at the whole pH range conditions. The importance of UO22+ lies in its potential incorporation of trace radioelements onto secondary uranyl phases. In view of the difficulty of U chemistry in natural groundwater, it is necessary to improve speciation assessment techniques so as to understand chemical processes. Raman spectroscopy has been shown as a powerful tool to analyze the speciation of various actinyl (UO22+,NpO2+ and PuO22+) and to determine the distribution of those elements which are more likely to be stable in a near-field groundwater environment. Therefore, the aim of this work is to follow UO22+ changes as a consequence of γ radiation in aqueous media under DGR conditions, and to understand the behavior of UO22+ as a function of aqueous media, which help to understand and predict the potential precipitation of the solid phases formed. In this work, the use of Raman spectroscopy adapted to the empirical analysis of different nuclear applications for initial uranium concentrations 0.04M at ambient atmosphere is shown, i.e. as monitoring tool for UO22+ precipitation as a function of pH, studying UO2(NO3)2·6H2O stability in aqueous solutions representative of groundwater, in particular at ionic strength I = 0.02 – 0.4 M and pH from 7 to 13.2; and to evaluate the effect of γ radiation fields. At 10−4-10-3 M of radiolytically formed H2O2 concentration, the amount of uranium in solution decreased, as a results of secondary phases precipitation. The results obtained will be useful to the performance assessment studies of the Spent Nuclear Fuel (SNF) stored in DGRs. The work performed provides a partial picture of secondary phase formations, as a result of corrosion of SNF in a DGR.

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.

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.

scholarly journals Study of Crater in the Gobi Desert Induced by Ground Explosion of Large Amounts of TNT Explosive up to 10 Tons

2021 ◽  
Vol 2021 ◽  
pp. 1-17
Author(s):  
Rongzheng Xu ◽  
Li Chen ◽  
Yuzhou Zheng ◽  
Zhan Li ◽  
Mingjin Cao ◽  
...  
Keyword(s):  
Conversion Coefficient ◽  
Near Field ◽  
Ground Surface ◽  
Failure Criteria ◽  
Gobi Desert ◽  
Large Weight ◽  
Fine Mesh ◽  
Contact Explosion ◽  
Protective Strategy ◽  
Sand And Gravel

Explosion craters on the ground surface induced by contact or near-field explosions have important implications, which can be used to assess blast consequences, guide the design of the explosion, or develop a protective strategy. In this study, to understand the crater characteristics induced by the contact explosion of large weight explosives, four field contact explosion tests were conducted on the surface of the Gobi Desert with large TNT charge weights of 1 ton, 3 tons, and 10 tons (test conducted twice). Cratering on the ground surface generated by large amounts of explosives was measured and evaluated, including the shape, depth, and diameter. A fine-mesh numerical model was developed and validated on the AUTODYN software platform, and a detailed parametric study was performed on the resulting craters. The effects of sand and gravel density, initiation method, shear modulus, and failure criteria were analyzed and discussed. An energy conversion coefficient was determined, and the corresponding theoretical equations were derived to predict the dimensions of the craters resulting from the large weight contact explosion. The calculated cratering characteristics were consistent with previous data and hence can be used in future engineering applications.

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