Isolation Of High-level Nuclear Waste Research Articles

Modeling gas breakthrough and flow phenomena through engineered barrier systems using a discrete fracture approach

Compacted clays are being considered to build engineered barrier systems (EBS) intended for the safe isolation of high-level nuclear waste (HLW) and spent nuclear fuel (SNF). The corrosion of the metallic canister containing the HLW/SNF will lead to the generation and buildup of the gas pressure in the more internal part of the clay buffer. This phenomenon would eventually trigger the formation and propagation of fractures in the clay barrier, jeopardizing its safety functions. In this work we propose to use the fragmentation technique (MFT) to model evolving fractures in clays triggered by gas pressurization. The MFT has been successfully used to model the formation of fractures in concrete, drying cracks in soil, hydraulic and thermo-fractures in rocks. In this work, we extend the MFT to deal with multiphase fluid flow in deformable porous media, and we upgraded a fully coupled computer finite element code using the extended technique. The proposed approach is first verified against analytical solutions and is then applied to model gas breakthrough experiments in clays. A very satisfactory performance of the method is observed in all the analyses, showing the potential of the MFT to tackle multiphase flow problems in deformable porous media with evolving discontinuities.

Corrosion of carbon steel in physically constrained locations in High level nuclear waste isolation systems

A method for calculating porosity of the corrosion product on the canister and of the engineered barriers (concrete) in the disposal of HLNW in Belgium supercontainer has been developed. The porosity of the rust is predicted to decrease substantially, such that the growth of the outer layer ceases and further corrosion is prevented. The decrease in the porosity of concrete is less than the decrease in the porosity of the rust. This is the consequence of the different thicknesses of the rust and concrete layers and different mechanical properties of the materials.

Lime mortar-compacted bentonite–magnetite interfaces: An experimental study focused on the understanding of the EBS long-term performance for high-level nuclear waste isolation DGR concept

The aim of this study was to obtain evidences regarding the physical and geochemical processes occurring as a result of the combined effects of cementitious materials from the concrete degradation and magnetite from steel corrosion on the bentonite barrier during disposal of high-level radioactive waste.A series of six experiments were done that attempt to reproduce the repository conditions prevailing from 1000 to 3000years after emplacement of wastes. A lime mortar was used as the source of calcium and alkalinity as this is the presumed reactive product produced during concrete degradation at long-term. Magnetite powder was used to simulate the final corrosion product of cast iron and C-steel under anaerobic conditions. Either a natural FEBEX bentonite or a pretreated “aged” sample, depleted in exchangeable Mg and enriched in K, were used as the swelling clay component. Experiments, with both types of bentonite, were performed simultaneously in cylindrical specimens (50mm diameter, 25mm length), confined in a Teflon® sleeve/steel case cells. These specimens were composed of cement mortar in contact with compacted bentonite, which was in turn in contact with compressed magnetite powder. They were hydrated with an artificial Na+–Ca2+×SO42−type Spanish reference clayey formation water for 18months at 60°C and constant hydraulic pressure applied through the base of the mortar.After dismantling and sampling the specimens, distribution of soluble ions, exchangeable cations and mineralogy were studied in the bentonite by different instrumental techniques. Iron migration or any impact of the corrosion products in the bentonite was not noticeable in the clay. Both, mortar and magnetite acted as sinks of chloride and sulfate. Small quantities of Ca–Al-sulfates and carboaluminates, which can allocate chlorides, were determined near the mortar-bentonite interface. Portlandite dissolved near the bentonite interface and induced the formation of calcium silicates hydrates (C-S-H) phases cementing the clay interface characterizing a calcium front that was developed from the mortar towards the bentonite. Magnesium silicate hydrates (M-S-H) phases were also concentrated at the interface with mortar in the natural bentonite. It was also determined that natural bentonite has potentially higher buffering capacity attenuating the calcium alkaline front than the pretreated clay. In both cases, a low porosity bentonite–mortar zone was experimentally created at the interface. This type of material should be carefully studied in order to predict the potential for further development of a diffusive alkaline alteration, the radionuclides retention and the consequences in the hydration rate of the unaffected bentonite buffer.

Plug Design for Deep Borehole Disposal of High-Level Nuclear Waste

This paper focuses on the improvement of the longevity and robustness of materials for sealing and plugging the upper portion of a deep borehole used for permanent isolation of high-level nuclear waste. Analytical models of porous and laminar flows show that even when materials have low intrinsic permeability, micron-sized cracks and gaps between the plug and rock (formed via chemical reaction, shrinkage, osmotic consolidation, etc.) significantly diminish the plug’s sealing properties. On this basis, materials such as asphalt, traditional cements, and pure bentonite—which crack or shrink under certain conditions—are unfavorable. An ongoing test program has formulated expanding cement mixtures containing MgO to prevent such bypass flow. Furthermore, these findings support using stable, malleable, and low-permeability plug material (k ≤ 10-16 m2), such as a crushed rock (70%) and bentonite (30%) mixture. Alternative clays such as sepiolite could be blended with the bentonite to further reduce the potential negative effects of salinity on bentonite permeability. A bounding and analytical model of a scenario where radionuclide escape is determined by advection through the plug (and assuming a large and constant driving pressure) shows that a plug permeability of 10-16 m2 is sufficiently low to prevent advective transport of radionuclides from a depth of 2 to 3 km to the surface within the timescale of interest (~1 million yr). Purely diffusive transport over the same distance, whether through the plug or host rock, is conservatively estimated to be significant only for a time >850 000 yr.

Passivity Breakdown on Copper: Influence of Chloride Ion

Passivity breakdown and pitting on pure copper, the assumed material for the outer canisters for the isolation of high-level nuclear waste in Sweden and Finland, has been studied in simulated, sulfide- and chloride-containing granitic rock groundwater and the data are interpreted in terms of the Point Defect Model (PDM). The near normal distribution in the critical breakdown potential (Vc) measured in deaerated sodium sulfide solutions with different chloride concentrations and for variable pH (= 8-10) is in satisfactory agreement with the quantitative prediction of the PDM. The critical areal concentration of condensed cation vacancies at the metal/film interface that leads to passivity breakdown, as predicted by the PDM from the experimental dependence of the Vc on the square root of the voltage sweep rate, ξ<1.09×1015cm−2, is in excellent agreement with that calculated from the unit cell dimensions of the substrate Cu (ξ ≈ 1.53×1015cm−2) and the barrier layer sulfide (Cu2S) (ξ ≈ 7.66×1014cm−2) for vacancy condensation upon the metal lattice or upon the cation sublattice of the film, respectively. This observation provides convincing evidence for the validity of the PDM for modeling passivity breakdown on Cu sulfide- and chloride-containing granitic rock groundwater.

Diffusion of soluble salts under a temperature gradient after the hydration of compacted bentonite

The transport of the soluble salts contained in a compacted bentonite has been studied in a hydration and heating experiment. The objective is to follow the processes of transport produced during and after the hydration of compacted bentonite in order to detect the generation of localized saline environments under the presence of a temperature gradient. Permanent saline environments may affect the surfacial and swelling properties of a bentonite when used as a clay liner for the isolation of high-level nuclear waste (HLNW). Several hermetic cells have been designed to constrain an unsaturated compacted bentonite cylinder (dry density of 1.65 g/cm 3). The bentonite is moisturized by injecting an aqueous solution in an opposite direction to a heat source. The tests are conducted during pre-saturation to post-saturation stages. The hydration is carried out with two different solutions, synthetic granitic water of low salinity and saline water. Two gradients with different average temperatures are generated: 100–60°C (16°C/cm) and 60–35°C (10°C/cm). Convection and diffusion processes govern the transport of salts in the compacted bentonite. The temperature is the variable that mostly influences the processes of transport, while the concentration of the injected solution had smaller importance. The convective transport of salts, driven by the hydration, occurs in the presence of the thermal gradient. By means of this mechanism, the dissolved ions concentrate on the heated area of the cell. The process permits the measurement of apparent diffusion coefficients ( D a) for ions and salts during the homogenization of concentration profiles in the post-saturation stages. D a for Cl −, SO 4 2−, Br −, Ca 2+, Mg 2+, Na +, K +, Sr 2+ and Zn 2+ have been calculated. The values vary in the range of 1.5×10 −10–4×10 −12 m 2/s; the ions Cl −, Na + and Ca 2+ being of higher mobility. Codiffusion of Ca 2+ and Na + with Cl − is the main mechanism found for salt transport in the compacted bentonite.

Modeling geological brines in salt-dome high level nuclear waste isolation repositories by artificial neural networks

In order to study the influence of various parameters on the acidity of simulated geological brines, an artificial intelligence technique based on neural network modeling has been developed. It has been found that the pH of simulated salt repository brines lies within the range of 3.2–5 as the temperature of the brine decays from 250° to 125°C. This environment might cause severe corrosion damage to canisters fabricated from carbon steel, particularly under slightly oxidizing conditions because of autocatalytical attack. It has also been demonstrated that artificial neural networks are efficient tools for analysing complex chemical systems, especially when conventional modeling is precluded by a lack of knowledge of the species and equilibria involved in the system.

Chemistry related to isolation of high-level nuclear waste

Chemistry related to isolation of high-level nuclear waste

An automated sensitivity analysis procedure for the performance assessment of nuclear waste isolation systems

The licensing of proposed high-level nuclear waste isolation facilities requires in-depth analysis of the expected performance and associated risks of the isolation system. Owing to the large uncertainties generally associated with the data characterizing geologic media and waste behavior and to the predictive nature of the simulations supporting the assessment, sensitivity and uncertainty analyses are necessary components of the overall performance assessment activities. Sensitivity analyses also play a major role in the identification of important parameters and of specific data needs for site characterization. To support an effort in making large-scale sensitivity analyses feasible, cost efficient and quantitatively complete, we have developed an automated procedure making use of computer calculus. The procedure, called GRESS (GRadient Enhanced Software System), is embodied in a precompiler that can process Fortran computer codes and add derivative-taking capabilities to the normal calculation scheme. In this paper, the automated GRESS procedure is described and applied to the code UCB-NE-10.2, which simulates the migration through a sorption medium of the radionuclide members of a decay chain. The sensitivity calculations for a sample problem are verified using comparison with analytical and perturbation analysis results. Conclusions are drawn relative to the applicability of GRESS for more general large-scale sensitivity studies, and the role of such techniques in an overall sensitivity and uncertainty analysis program is discussed.

Thermal expansion and cracking of three confined water-saturated igneous rocks to 800�C

Solutions of engineering problems of very deep drilling, geothermal energy production, and high-level nuclear-waste isolation require adequate understanding of the mechanical and transport properties of rocks at relatively low pressures but high temperatures. Accordingly, the thermal expansions of water-saturated Charcoal Granite, Mt. Hood Andesite, and Cuerbio Basalt have been measured at effective confining pressures (P e ) of 5, 50, and 100 MPa to 800° C. The mean coefficient of linear thermal expansion (α) is a function of lithology,P e , temperature (T) and initial porosity (ϕ). For example, for the Charcoal Granite, α increases withT at all pressures. The signature of the alpha-beta transition of quartz is more pronounced at the lower pressures; at 100 MPa α nearly mimics that of a crack-free rock forT<300° C. α for the andesite atP e =5 MPa ranges from 10 to 15×10−6/°C from 200° to 400° C then decreases gradually to 10.1×10−6/°C at 800° C. At 50 MPa α ranges from 11.7×10−6/°C at 100° C to 8.6×10−6 at 200°C, then increases at a much lower rate to 11×10−6 at 600° C. The basalt, however, has an essentially constant α (11×10−6/°C) forT>150°C at the lower pressure and shows but a small increase in α from 6 to 9×10−6 from 100° to 800° C at 50 MPa. The difference between measured values of thermal expansion and those calculated from simple mixture-theory relates to new crack porosity generated as a result of differential thermal expansion at the anisotropic grain scale. For the granite, a two to three order of magnitude increase in permeability (k) is predicted from the relation,k∝φ 3.

Selection of Barrier Metals for a Waste Package in Tuff

ABSTRACTThe Nevada Nuclear Waste Storage Investigations (NNWSI) project under the Civilian Radioactive Waste Management Program is planning a repository at Yucca Mountain at the Nevada Test Site for isolation of high-level nuclear waste. Lawrence Livermore National Laboratory is developing designs for an engineered barrier system containing several barriers such as the waste form, a canister and/or an overpack, packing, and near field host rock. In this paper we address the selection of metal containment barriers.

Payments In-Lieu-Of Taxes Between Federal and Local Governments and the Siting of Federally-Owned Waste Isolation Activities

This paper considers the costs that accrue to a local economy when a high level nuclear waste isolation facility is site. To accept a nuclear waste site, the community must be compensated in some way. Payments-in-lieu-of-taxes results from the tax exempt nature of the federal facility and are justified on the grounds that the federal government has a responsibility to pay its own way when it acts as a public marketeer. This paper presents an analysis of transfer payment mechanisms that could address these cost issues.