Modelling groundwater contamination above a nuclear waste repository at Gorleben, Germany

The candidate repository for high-level nuclear waste in the Gorleben salt dome, Germany, is expected to host 8,550 tonnes of uranium in burnt fuel. It has been proposed that 5,440 waste containers be deposited at a depth of about 800?m. There is 260?C280?m of siliciclastic cover sediments above the proposed repository. The potential groundwater contamination in the siliciclastic aquifer is simulated with the TOUGHREACT and TOUGH2-MP codes for a three-dimensional model with 290,435 elements. Two deterministic cases are simulated. The single-phase case considers the transport of radionuclides in the liquid phase only. The two-phase case accounts for hydrogen gas generated by the corrosion of waste containers and release of gaseous C-14. The gas release via a backfilled shaft is assumed to be steady (non-explosive). The simulation period is 2,000,000 years for the single-phase case and 7,000 years for the two-phase case. Only the radioactive dose in the two-phase case is higher than the regulatory limit (0.1?mSv/a).     

Modelling radionuclide transport in large fractured-media systems: the example of Forsmark, Sweden

The planned high-level nuclear waste repository at Forsmark, Sweden, will accommodate 6,824 containers with a total of 13,920 tonnes of uranium in burnt fuel at approximately 400 m depth in a fractured-granite aquifer. The transport of radionuclides, which may be released from the disposed waste, is simulated with the TOUGHREACT code for a three-dimensional model with 305,571 elements. The model performs coupled flow-transport simulations. It aims to achieve more realistic simulations of contaminant transport than the commonly used decoupled procedure consisting of three-dimensional flow and one-dimensional transport simulations. The model has a relatively small problem size because it is designed as a double-porosity model (one matrix continuum) that is the parameterised equivalent of a much larger multiple-interacting continua (MINC) model, i.e. a model with a finely discretised matrix (several matrix continua). The parameterisation is performed with two-dimensional models. Only one or two variables among three variables (diffusive transport distance between fracture and matrix, retardation factor and effective diffusivity) have to be parameterised. The results obtained with the parameterised three-dimensional model are very close to those that can be obtained with a much larger MINC model but may be quite different from those that can be obtained with the conventional decoupled procedure.     

Experimental determination of the speciation,partitioning, and release of perrhenate as a chemical surrogate for pertechnetate from a sodalite-bearing multiphase ceramic waste form

A key component to closing the nuclear fuel cycle is the storage and disposition of nuclear waste in geologic systems. Multiphase ceramic waste forms have been studied extensively as a potential host matrix for nuclear waste. Understanding the speciation, partitioning, and release behavior of radionuclides immobilized in multiphase ceramic waste forms is a critical aspect of developing the scientific and technical basis for nuclear waste management. In this study, we evaluated a sodalite-bearing multiphase ceramic waste form (i.e., fluidized-bed steam reform sodium aluminosilicate [FBSR NAS] product) as a potential host matrix for long-lived radionuclides, such as technetium (⁹⁹Tc). The FBSR NAS material consists primarily of nepheline (ideally NaAlSiO₄), anion-bearing sodalites (ideally M₈[Al₆Si₆O₂₄]X₂, where M refers to alkali and alkaline earth cations and X refers to monovalent anions), and nosean (ideally Na₈[AlSiO₄]₆SO₄). Bulk X-ray absorption fine structure analysis of the multiphase ceramic waste form, suggest rhenium (Re) is in the Re(VII) oxidation state and has partitioned to a Re-bearing sodalite phase (most likely a perrhenate sodalite Na₈[Al₆Si₆O₂₄](ReO₄)₂). Rhenium was added as a chemical surrogate for ⁹⁹Tc during the FBSR NAS synthesis process. The weathering behavior of the FBSR NAS material was evaluated under hydraulically unsaturated conditions with deionized water at 90 °C. The steady-state Al, Na, and Si concentrations suggests the weathering mechanisms are consistent with what has been observed for other aluminosilicate minerals and include a combination of ion exchange, network hydrolysis, and the formation of an enriched-silica surface layer or phase. The steady-state S and Re concentrations are within an order of magnitude of the nosean and perrhenate sodalite solubility, respectively. The order of magnitude difference between the observed and predicted concentration for Re and S may be associated with the fact that the anion-bearing sodalites contained in the multiphase ceramic matrix are present as mixed-anion sodalite phases. These results suggest the multiphase FBSR NAS material may be a viable host matrix for long-lived, highly mobilie radionuclides which is a critical aspect in the management of nuclear waste.     

The COUPLEX Test Cases: Nuclear Waste Disposal Simulation

The models appearing in the COUPLEX benchmark are a set of simplified albeit realistic test cases aimed at simulating the transport of radionuclides around a nuclear waste repository. Three different models were used. The first test case is related to simulations based on a simplified 2D far-field model close to those used for safety assessments in nuclear waste management. It leads to a classical convection diffusion type problem, but with highly variable parameters in space, highly concentrated sources in space and time, very different time scales and accurate results expected even after millions of years. The second test case is a simplification of a typical 3D near-field computation, taking into account the glass dissolution of vitrified waste, and the congruent release of several radionuclides (including daughter products), with their migration through the geological barrier. The aim of the third test case is to use the results of the near-field computation (COUPLEX 2) to drive the behavior of the nuclide source term in the far-field computation (COUPLEX 1). The modeling of this last case was purposely left rather open, unlike the previous two, leaving the choice to participants of the way the coupling should be made.     

Geochemical assessement of actinide isolation in a German salt repository environment

A significant criterion in evaluating disposal strategies for high-level nuclear waste is the assessment of the isolation capacity for the most radiotoxic radionuclides, the actinides. Important processes pertinent to potential mobilization from the waste forms, retention in secondary phases and migration of actinides in the geochemical environment of the near field of disposal locations are summarized. Criteria are formulated for assessing engineered barrier performance as a geochemical barrier for actinide long-term retention.     

A Continuum One——Dimensional Model for Radionuclide Migration Testing in Unsaturated Loess

Unsaturated radionuclide migration experiments were conducted in a pit inside thetesting hall.Several types of radionuclides were used in the experiments.Tritium wasused as a tracer for water movement in unsaturated loess.Other kinds of radionuclideswere also used in order to obtain fundamental parameters for radionuclide migration sothat further environmental assessment of low—level radioactive waste disposal can be car-ried out.Mechanisms governing unsaturated flow in loess,that is,principles ofone—way lateral flow,are presented qualitatively in this paper.And a continuumone—dimensional model for radionuclide migration testing is developed based on the ex-periments conducted under the particular conditions at the test site.The data measuredfrom the tests were compared with solutions of this one—dimensional model.Resultsshow that this model is feasible for modeling radionuclide migration in unsaturatedloess.     

Near-field high temperature transport: evidence from the genesis of the Osamu Utsumi uranium mine, Poços de Caldas alkaline complex, Brazil

The chemical, isotopic and mineralogical alteration which occurred during primary uranium ore deposition at the breccia pipe-hosted Osamu Utsumi mine, Poços de Caldas, Brazil was studied as a natural analogue for near field radionuclide migration. Chemical and isotopic alteration models were combined with finite difference models of the convective cooling of caldera intrusives. The modeling indicates that the intense chemical, isotopic, and mineralogical alteration of the Osamu Utsumi breccia pipe requires the circulation of > 10⁵ kg/cm² of boiling hydrothermal fluid > 200°C through each square centimeter cross-section of the pipe. This circulation could be driven by heat from a 6 km diameter intrusive extending to 10 km depth. Even with this large amount of circulation concentrated in the permeable breccia pipe, uranium solubilities must be orders of magnitude greater than indicated in the most recent experiments (and more in line with previous estimates) to produce the primary uranium mineralization at the Osamu Utsumi mine.The same models applied to a hypothetical high temperature waste repository show that heat from radioactive decay will produce a hydrothermal circulation system remarkably similar to that studied at the natural analogue site at Poços de Caldas. The depth of fluid convection induced by the hypothetical repository would be 5 to 10 km, the maximum temperature would be 300°C, the lifetime of the high temperature phase would be a few thousand years, and boiling would occur and cause most of the alteration within the hypothetical waste repository. This physical analysis emphasizes the importance of permeability on a 10 × 10 × 10 km scale in controlling the potential amount of circulation through the hypothetical repository.Application of the chemical models successfully used to interpret mineralization and alteration at the Poços de Caldas Osamu Utsumi mine to the hypothetical waste repository shows that even in a worst case scenario (waste implaced in a permeable host rock with no measures taken to inhibit flow though the repository) the amount of hydrothermal alteration in the hypothetical repository will be 0.1% of that in the breccia pipe at Osamu Utsumi. Assuming no barriers to uranium mobility, uranium precipitation above the hypothetical repository would be 0.04 ppm (rather than 40 ppm), hydrothermal alteration 0.03 wt% (rather than 30 wt%), etc.Our analysis indicates that modeled mineralogical alteration is sensitive to the thermodynamic data base used. Prediction of mineralogical alteration (which may be necessary to predict the migration of radionuclides other than uranium, for example) probably cannot be based directly on even very carefully collected laboratory thermodynamic data. Mineralogical complexities of the system, as well as data base uncertainties will require calibration of the thermodynamic framework against mineralogical alteration observed in the laboratory or field.     

Confinement matrices for low- and intermediate-level radioactive waste

Mining of uranium for nuclear fuel production inevitably leads to the exhaustion of natural uranium resources and an increase in market price of uranium. As an alternative, it is possible to provide nuclear power plants with reprocessed spent nuclear fuel (SNF), which retains 90% of its energy resource. The main obstacle to this solution is related to the formation in the course of the reprocessing of SNF of a large volume of liquid waste, and the necessity to concentrate, solidify, and dispose of this waste. Radioactive waste is classified into three categories: low-, intermediate-, and high-level (LLW, ILW, and HLW); 95, 4.4, and 0.6% of the total waste are LLW, ILW, and HLW, respectively. Despite its small relative volume, the radioactivity of HLW is approximately equal to the combined radioactivity of LLW + ILW (LILW). The main hazard of HLW is related to its extremely high radioactivity, the occurrence of long-living radionuclides, heat release, and the necessity to confine HLW for an effectively unlimited time period. The problems of handling LILW are caused by the enormous volume of such waste. The available technology for LILW confinement is considered, and conclusion is drawn that its concentration, vitrification, and disposal in shallow-seated repositories is a necessary condition of large-scale reprocessing of SNF derived from VVER-1000 reactors. The significantly reduced volume of the vitrified LILW and its very low dissolution rate at low temperatures makes borosilicate glass an ideal confinement matrix for immobilization of LILW. At the same time, the high corrosion rate of the glass matrix at elevated temperatures casts doubt on its efficient use for immobilization of heat-releasing HLW. The higher cost of LILW vitrification compared to cementation and bitumen impregnation is compensated for by reduced expenditure for construction of additional engineering barriers, as well as by substantial decrease in LLW and ILW volume, localization of shallow-seated repositories in various geological media, and the use of inexpensive borosilicate glass.     

Thermal and physical properties of reconsolidated crushed rock salt as a function of porosity and temperature

Crushed salt can be used as backfill to bury and conduct heat away from radioactive waste in salt repositories. As the crushed salt compacts during reconsolidation, its thermal, mechanical and hydrologic properties will change in a manner related to the porosity. Measurements of crushed salt thermal properties are conducted to evaluate such relationships. A simple mixture theory model is presented to predict thermal conductivity of consolidating salt in repository conditions. Experimental work was completed to evaluate the model by measuring thermal conductivity, thermal diffusivity and specific heat of crushed salt as a function of porosity and temperature. Sample porosity ranged from 0 to 46 %, and measurements were made at ambient pressure, from room temperature to 300 °C. These are the temperature conditions expected in a radioactive waste storage facility. Crushed salt thermal conductivity decreases with increasing porosity and temperature. Thermal diffusivity showed little porosity dependence but decreased with increasing temperature. Specific heat also shows little porosity dependence but increases with increasing temperature. Fracture porosity in deformed bedded salt cores appears to reduce thermal conductivity more dramatically than inter- and intra-granular porosity in consolidated salt. A long-term effort to dry crushed salt at high temperatures resulted in a 0.48 weight-percent loss of water that had resided at grain boundaries and in intra-granular fluid inclusions. While this loss does not significantly affect thermal properties, the release of this water volume could impact the mechanical response of the reconsolidating salt and host rock.     

Model of colloidal transportation of radionuclides by groundwater

The transportation of colloidal radionuclides by groundwater was subject to theoretical analysis. The far field of radioactive contamination of the underground environment (liquid waste pumping sites or storage of solidified waste) is dominated by pseudocolloids, i.e., colloidal particles of natural origin contaminated with radionuclides upon contact of groundwater with radioactive materials. Properties of real pseudocolloids were analyzed at sites of radioactive contamination of the underground environment. Based on a probabilistic approach, we developed a mathematical model of pseudocolloid transportation by groundwater, taking into account the difference in size of colloidal particles and the occurrence of nonradioactive natural particles with a similar composition in the groundwater. It is proved that nonuniform dimensions of the particles considerably affect the water transportation rate.     

Natural analogues

Natural analogues, which are occurrences of materials or processes which are analogous to expected materials or processes in a waste repository, are used to clarify the ability of the geologic environment and engineered system to contain the waste. It is hoped that natural analogues can provide data and insight on processes which generally occur on temporal or geometric scales which are too large to be studied in laboratory or field experiments. Natural analogues have been described for the study of transport and migration of radionuclides through the backfill and host rock, and the stability of backfill, shaft seals, waste forms, and waste containers. Natural analogues have also been suggested as an aid in predicting overall repository performance.     

Micro-spectroscopic investigations of highly heterogeneous waste repository materials

Assuring safe disposal and long-term storage of radioactive and toxic wastes corresponds to a primary environmental task of present societies. To improve any technical limitation, a mechanistic understanding of the processes governing the binding of heavy metals and radionuclides is required. In this study, the significance of synchrotron-based X-ray microprobes for elucidating the spatial distribution and the speciation of radionuclides in highly heterogeneous waste repository materials will be outlined. A case study on the uptake process of Co in cementitious engineered barrier materials exposed to microbial degradation will be presented.     

Adsorption of Co and selected actinides by Mn and Fe oxides in soils and sediments

Iron and Mn oxides and associated radionuclides in soils and sediments from the radioactive waste burial grounds at Oak Ridge National Laboratory have been selectively extracted using wet chemical techniques. Product-moment-correlation analyses have demonstrated that ⁶⁰Co and various actinides, principally ²⁴⁴Cm, ²⁴¹Am and ²³⁸Pu are dominantly associated with Mn oxides. Correlation coefficients between these radionuclides and Fe oxides and organic C are generally very low. The important role of Mn oxides in radionuclide adsorption is attributed to their unique surface and colloidal properties. The data illustrate the importance of the Mn oxide component of soils and sediments in controlling transition metal and actinide solubility.These results suggest two major implications for the disposal of radioactive waste. First, in order to minimize future ⁶⁰Co and actinide mobilization from disposal sites, a chemical environment in which Mn oxides are least soluble should be maintained. Second, the liberal use of Mn oxides in waste management operations might improve long-term retention of these radionuclides. Deep-sea Mn modules, which may in the future be mined for their trace metal contents, could serve as a ready supply of Mn oxide for waste disposal applications.     

Preoperational assessment of solute release from waste rock at proposed mining operations

Environmental assessments are conducted prior to mineral development at proposed mining operations. Among the objectives of these assessments is prediction of solute release from mine wastes projected to be generated by the proposed mining and associated operations. This paper provides guidance to those engaged in these assessments and, in more detail, provides insights on solid-phase characterization and application of kinetic test results for predicting solute release from waste rock. The logic guiding the process is consistent with general model construction practices and recent publications. Baseline conditions at the proposed site are determined and a detailed operational plan is developed and imposed upon the site. Block modeling of the mine geology is conducted to identify the mineral assemblages present, their masses and compositional variations. This information is used to select samples, representative of waste rock to be generated, that will be analyzed and tested to describe characteristics influencing waste rock drainage quality. The characterization results are used to select samples for laboratory dissolution testing (kinetic tests). These tests provide empirical data on dissolution of the various mineral assemblages present as waste rock. The data generated are used, in conjunction with environmental conditions, the proposed method of mine waste storage, and scientific and technical principles, to estimate solute release rates for the operational scale waste rock.Common concerns regarding waste rock are generation of acidic drainage and release of heavy metals and sulfate. Key solid phases in the assessments are those that dissolve to release acid and sulfate (iron sulfides, soluble iron sulfates, hydrated iron-sulfate minerals, minerals of the alunite–jarosite group), those that dissolve to neutralize acid (calcium and magnesium carbonates, silicate minerals), and those that release trace metals (trace metal sulfides, hydrated trace metal-sulfate minerals). Conventional mineralogic, petrographic, and geochemical analyses generally can be used to determine the quantities of these minerals present and to describe characteristics that influence their dissolution. A key solid-phase characteristic is the mineral surface area exposed for reaction, which is influenced by mode of occurrence (included, interstitial, liberated) and the extent of mineral surface coating. Short-term dissolution tests can estimate the extent of hydrated sulfate minerals present. Longer term dissolution tests are necessary to describe the dependence of drainage pH and solute release rates on solid-phase variation. The extensive data compiled from baseline pre-development definition, the operational plan, solid-phase characterization, and dissolution testing are ultimately synthesized by means of a modeling exercise requiring considerable technical and scientific expertise. The predicted rates (model outputs) are expressed as probability distributions to allow assessment of risk. This exercise must be technically defensible and transparent so that regulators can confidently assess the results and evaluate the operational plan proposed. Technical and non-technical challenges involved in implementing such programs are identified to benefit management planning for both industry and government.     

The French experience in low level radioactive waste management

The Radioactive Waste Management Agency (ANDRA-FRANCE) is now operating a new facility in the eastern part of the Paris basin which is designed to dispose of one million cubic meters of waste.

The safety of the waste disposal is based on a multibarrier concept including waste packages, concrete disposal modules, site and closure operations.

Under normal conditions, confinement is guaranteed by the waste packages and the disposal modules, as they prevent the waste from being leached by rainfall or underground water over a certain period of time.

The site must bring an additional guarantee concerning the isolation of waste from water. Consequently, the chosen site must be located in an area where no natural disasters (landslides, earthquakes, etc.) can harm the isolating barriers. The geological, hydrogeological and chemical characteristics must allow us to minimize and control the transport of radionuclides within the ground. Finally, the chosen site must be in an area where it is easy to implement a system to monitor the environment.

A set of criteria guides the choice of site. The criteria include such factors as low seismicity, geotechnical stability, a hydrogeology that is simple to model, a location sufficiently above the water table and safe from the threat of flooding, good radionuclide sorption and the absence of any mineral or other natural resources of economic interest.

At the time of the closure of the disposal facility, the entire collection of modules will be covered by an impervious cap composed of clayey layers interbedded by sandy layers and overlain by humus to promote the growth of grass. The facility will then look like a succession of undulating green mounds.

A 300-year monitoring period will follow the closure. During this period, the water collecting networks and cap will be maintained and radioactivity in underground and stream water will be controlled.

We have selected the AUBE site as a case study to illustrate the French waste management experience. We will report on how the site characterization program has been calTied out, including the hydrogeological modelling which is being applied to both the operating and post-closure periods.     

深井地下灌注数值模型SWIFT

深井地下灌注是一种有效处理工业废液的方法,随着认识的不断深入和技术的不断提高,将成为未来我国对难处理、高毒性废液实行最终处理的重要选择之一。SWIFT数值模型能够同时模拟灌注层中水流运动、热量运移、盐分运移和放射性核素运移过程,并且能够充分考虑深部灌注层高温高压等特殊地质环境。其多方面的模拟功能不但适用于深井灌注的数值模拟,还可用于海水入侵、高放射性核废料填埋的模拟分析。在美国,该数值模型已经被用于多个深井灌注项目的数值模拟。对SWIFT进行了较全面的介绍,并指出该模型使用过程中应特别注意的几个问题。     

Controls on permafrost thaw in a coupled groundwater-flow and heat-transport system: Iqaluit Airport,Nunavut, Canada

Numerical simulations of groundwater flow and heat transport are used to provide insight into the interaction between shallow groundwater flow and thermal dynamics related to permafrost thaw and thaw settlement at the Iqaluit Airport taxiway, Nunavut, Canada. A conceptual model is first developed for the site and a corresponding two-dimensional numerical model is calibrated to the observed ground temperatures. Future climate-warming impacts on the thermal regime and flow system are then simulated based on climate scenarios proposed by the Intergovernmental Panel on Climate Change (IPCC). Under climate warming, surface snow cover is identified as the leading factor affecting permafrost degradation, including its role in increasing the sensitivity of permafrost degradation to changes in various hydrogeological factors. In this case, advective heat transport plays a relatively minor, but non-negligible, role compared to conductive heat transport, due to the significant extent of low-permeability soil close to surface. Conductive heat transport, which is strongly affected by the surface snow layer, controls the release of unfrozen water and the depth of the active layer as well as the magnitude of thaw settlement and frost heave. Under the warmest climate-warming scenario with an average annual temperature increase of 3.23 °C for the period of 2011–2100, the simulations suggest that the maximum depth of the active layer will increase from 2 m in 2012 to 8.8 m in 2100 and, over the same time period, thaw settlement along the airport taxiway will increase from 0.11 m to at least 0.17 m.