Hypothetical Repository Research Articles

The value of simplified models of radionuclide transport for the safety assessment of nuclear waste repositories: A benchmark study

In order to assess sites for a deep geological repository for storing high-level nuclear waste safely in Germany, various numerical models and tools will be in use. For their interaction within one workflow, their reproducibility, and reliability version-controlled open-source solutions and careful documentation of model setups, results and verifications are of special value. However, spatially fully resolved models including all relevant physical and chemical processes are neither computationally feasible for large domains nor is the data typically available to parameterize such models. Thus, simplified models are crucial for the pre-assessment of possible sites to narrow down the list of suitable candidates for which detailed site investigations and fully resolved models will be done at a later stage. Still, the accuracy of these simplified models is of importance as the pre-assessment of suitable sites will be based on them. In this study, we compare the modelling capabilities of TransPyREnd, a one-dimensional transport code based on finite differences, specifically developed for the fast estimation of radionuclide transport by the German federal company for radioactive waste disposal (BGE), with OpenGeoSys, which is a modelling platform based on finite elements in up to three spatial dimensions. Both codes are used in the site selection procedure for the German nuclear waste repository. The comparison of the model results of TransPyREnd and OpenGeoSys is augmented by comparisons with an analytical solution for a homogeneous material. For the purpose of numerical benchmarking, we consider a geological profile located in southern Germany as an example where the hypothetical repository is located in a clay-stone formation. TransPyREnd and OpenGeoSys yield overall similar results. However, both codes use different discretizations which impact is the highest for strongly sorbing compounds, while the difference gets negligible for less sorbing and more diffusive compounds as higher diffusion tends to blur the initial conditions. Overall, the OpenGeoSys model is more exact whereas the TransPyREnd model has considerable faster run times. We found in our example, that significant substance amounts are only leaving the host rock formation, if apparent diffusion is high, for which case both codes give similar results, while relative differences are considerable for strongly sorbing compounds. However, in the latter case no significant substance amount of radionuclides leaves the host-rock formation, thus deeming the differences in the model results minor for the overall safety assessments of sites.

Modelling of Radiological Health Risk in Water from Abstraction Well Close to a Hypothetical Radioactive Waste Repository in the Accra Plains

This study evaluated the integrity of the geosphere of the Accra Plains to host a radioactive waste disposal facility for Ghana’s radioactive waste materials. The study thus assessed the migration of radioactive contaminant from a hypothetical radioactive waste disposal facility through the geosphere to the biosphere in the Accra Plains. The Accra Plains is predominantly underlain by schist rocks, hence the hydrogeologic parameters of schist rock from literature were used as input data. AMBER software was used to develop a model to estimate the peak total annual effective dose from all the radionuclides at various receptor locations. The endpoint of interest was the dose received by a receptor for ingesting water from an abstraction well located at distance; 100 m, 200 m, 500 m, 1 km, 2 km, 3 km, 4 km and 5 km from the hypothetical repository. From the simulation, the annual effective dose values ranged from 2.07E-25 Sv/y to 6.45E-17 Sv/y which are all below the International Commission on Radiological Protection (ICRP) dose constraint of 0.3 mSv/y. The lifetime cancer risk for ingesting water from the abstraction wells were then calculated from the annual effective dose values. The resulting values ranged from 1.06E-24 to 3.30E-16 thus indicating that the potential cancer risk for ingesting water from the abstraction wells located within 100 m to 5 km from the hypothetical disposal facility is negligible. The model results demonstrate that, siting a disposal facility in the study area might not pose any significant risk to the sprawling population and the environment.

Gas transport modelling at different spatial scales of a geological repository in clay host rock

The work presented in this article is a part of the international European Commission FORGE project which studied the behaviour of repository gases in the context of radioactive waste disposal. Experimental work is essential to understand the main transport mechanisms for gas and to determine the main geophysical parameters. However, while laboratory and rock experiments can help to investigate the short- and partly the middle-term behaviour of gas in a rock formation, long-term predictions have to be based on numerical simulations. Numerical simulation of long-term gas behaviour in a deep geological repository was one of the aims in the FORGE project. The objectives of the FORGE modelling were to describe the state-of-the-art consideration of gas in performance assessment, and to propose an updated treatment of gas issues in long-term safety assessments for high-level radioactive waste and spent nuclear fuel. Three benchmark exercises for a hypothetical geological repository in clay host rock ranging in scale from a single disposal cell to the whole repository were defined. To provide added value to this benchmark, a feature not yet well represented in typical gas simulations was introduced: the explicit representation of the interfaces between the excavation disturbed zone and the engineered elements within the excavation, such as waste canisters, bentonite plugs, and access drifts. In order to assess gas transport behaviour at the whole repository scale, models were developed with ‘mathematical’ or ‘numerical’ upscaling techniques for small-scale features. The assessment across different modelling scales revealed the main long-term gas migration pathways and led to the conclusion that the explicit representation of interfaces is not important.

Preliminary analysis of gaseous radiocarbon behavior in a geological repository hosted in salt rock

ABSTRACTA preliminary evaluation of gaseous radiocarbon (14C) behavior under geological repository conditions for Italian radioactive high level waste-long-lived and intermediate level waste disposal has been performed. Although in Italy there is still no defined project for a geological disposal facility, current work may support future safety assessment studies for a hypothetical future repository in deep salt rock. In the Italian context of radioactive waste, the percentage of 14C bearing waste to be disposed in a possible geological repository is low; irradiated graphite is the most important radiological source. Data about the radiological inventory has been collected to simulate production and migration of gaseous 14C in a hypothetical geological repository. Three different conceptual models have been developed and simulated. The first model has considered a preliminary evaluation of the radiological impact referred to the whole inventory; the second and third model have evaluated the impact only due to the irradiated graphite. A preliminary sensitivity analysis has been carried out, highlighting the importance of geometry and of distribution coefficients (Kd) in materials used to seal the disposal underground facility. Results show the possibility to correlate the Kd values, the volume and the location of the sealing materials to the amount of 14C migrating toward the surface.

A comparative study of discrete fracture network and equivalent continuum models for simulating flow and transport in the far field of a hypothetical nuclear waste repository in crystalline host rock

One of the major challenges of simulating flow and transport in the far field of a geologic repository in crystalline host rock is related to reproducing the properties of the fracture network over the large volume of rock with sparse fracture characterization data. Various approaches have been developed to simulate flow and transport through the fractured rock. The approaches can be broadly divided into Discrete Fracture Network (DFN) and Equivalent Continuum Model (ECM). The DFN explicitly represents individual fractures, while the ECM uses fracture properties to determine equivalent continuum parameters. We compare DFN and ECM in terms of upscaled observed transport properties through generic fracture networks. The major effort was directed on making the DFN and ECM approaches similar in their conceptual representations. This allows for separating differences related to the interpretation of the test conditions and parameters from the differences between the DFN and ECM approaches. The two models are compared using a benchmark test problem that is constructed to represent the far field (1×1×1km3) of a hypothetical repository in fractured crystalline rock. The test problem setting uses generic fracture properties that can be expected in crystalline rocks. The models are compared in terms of the: 1) effective permeability of the domain, and 2) nonreactive solute breakthrough curves through the domain. The principal differences between the models are mesh size, network connectivity, matrix diffusion and anisotropy. We demonstrate how these differences affect the flow and transport. We identify the factors that should be taken in consideration when selecting an approach most suitable for the site-specific conditions.

Modelling groundwater flow changes due to thermal effects of radioactive waste disposal at a hypothetical repository site near Sellafield, UK

High-level radioactive waste (HLW) and spent fuel from nuclear power plants (SF) require very long duration disposal because of its long-lived toxicity. Specific repository design is necessary to accommodate the associated large heat generation from continual decay. Conceptual designs propose geological containment within salt, claystone, or crystalline bedrock. However, while many studies have investigated the safety of disposal in these host rocks, the interaction of near-field heat coupled to far-field groundwater flow for UK-relevant conditions has not been fully investigated. Using an excellent dataset from a repository investigation in saturated fractured rock at Sellafield, UK, with deep groundwater flow, a preliminary investigation of the effect of the addition of heat from HLW and SF on regional groundwater flow is undertaken. A coupled thermo-hydrogeological groundwater flow model dependent on fluid density and viscosity has been calibrated to field data, before heat is added to simulate waste. Substantial heating reduces fluid density and dynamic viscosity. Groundwater flows upwards through permeable damage zones of high-angle faults and increases natural velocity by 1.5×. Elevated groundwater temperatures are predicted at the surface in just hundreds of years. Travel times for conservative chemical species from the hypothetical repository to the St. Bees Sandstone aquifer are reduced by over one-third if engineered barriers fail at the point of containment, and remain reduced by hundreds of years for particles released thousands of years after containment. The results imply that in conditions similar to those modelled, the high-performance guarantee of the engineered barrier is essential to repository safety. A very good understanding of the coupled interaction of near-field heat and far-field groundwater flow will be important for safety cases, especially accuracy concerning heterogeneous fault discontinuities. It is necessary to model heat advection and faults at fractured sites.

FASTREACT – An efficient numerical framework for the solution of reactive transport problems

In the framework of safety assessment studies for geological disposal, large scale reactive transport models are powerful inter-disciplinary tools aiming at supporting regulatory decision making as well as providing input to repository engineering activities. Important aspects of these kinds of models are their often very large temporal and spatial modelling scales and the need to integrate different non-linear processes (e.g., mineral dissolution and precipitation, adsorption and desorption, microbial reactions and redox transformations). It turns out that these types of models may be computationally highly demanding. In this work, we present a Lagrangian-based framework, denoted as FASTREACT, that aims at solving multi-component-reactive transport problems with a computationally efficient approach allowing complex modelling problems to be solved in large spatial and temporal scales. The tool has been applied to simulate radionuclide migration in a synthetic heterogeneous transmissivity field and the results have been successfully compared with those obtained using a standard Eulerian approach. Finally, the same geochemical model has been coupled to an ensemble of realistic three-dimensional transport pathways to simulate the migration of a set of radionuclides from a hypothetical repository for spent nuclear fuel to the surface. The results of this modelling exercise, which includes key processes such as the exchange of mass between the conductive fractures and the matrix, show that FASTREACT can efficiently solve large-scale reactive transport models.

KURT 환경 자료를 이용한 가상의 다중 발생원에서의 누출 핵종의 이동 시간 평가

KURT(KAERI Underground Research Tunnel) 부지 부근에 가상의 처분장을 설정하고, 해당 부지의 세 지점에서 방사성폐기물로부터 누출된다고 가정한 방사성 핵종의 이동 시간을 계산하였다. 핵종의 이동 경로는 핵종 누출 지점에서 천부 지하수대까지로 설정하고 KURT 주변 지하수 유동계 모의를 통해 결정하였다. 세 지점은 지하수가 빠르게 유동하는 구조(highly water-conductive feature)를 지나가기 때문에 천부 지하수까지 도달하는데 상대적으로 적은 시간이 걸리는 지점으로 선정되었다. 핵종의 이동 시간은 TDRW(Time-Domain Random Walk) 기법을 통해 계산하였다. 지하수 내의 핵종의 이동 시간을 계산하기 위해, 이류(advection)와 분산(dispersion) 이외에 암반 기질(rock matrix)로의 확산(diffusion)과 기질 내부에서의 흡착(sorption)이 고려되었고, 핵종의 붕괴 및 변환에 의한 영향도 몇 개의 붕괴 사슬(decay chain)을 이용하여 계산에 반영하였다. 계산 결과를 보면, 지표 부근의 천부 지하수에 도달하는 핵종의 시간당 이동량(mass flux)은 복수의 이동 경로뿐만 아니라 핵종의 반감기와 암반 기질 내에서의 핵종의 흡착 분배 계수에 크게 영향을 받는 것으로 나타났다. 따라서 보다 안정적이고 불확실성이 감소된 심지층 처분장의 안전성 평가를 위해 우선적으로 필요한 사항으로는, 장반감기 핵종에 대한 평가가 이동 과정 이외에 저장 용기에 들어있는 상태에서부터 면밀하게 이루어져야 하고, 암반 기질에서 발생하는 핵종의 흡착 과정이 심부 현장 조건을 반영하여 평가되어야 할 것으로 생각된다. A hypothetical repository was assumed to be located at the KURT (KAERI Underground Research Tunnel) site, and the travel times of radionuclides released from three source positions were calculated. The groundwater flow around the KURT site was simulated and the groundwater pathways from the hypothetical source positions to the shallow groundwater were identified. Of the pathways, three pathways were selected because they had highly water-conductive features. The transport travel times of the radionuclides were calculated by a TDRW (Time-Domain Random Walk) method. Diffusion and sorption mechanisms in a host rock matrix as well as advection-dispersion mechanisms under the KURT field condition were considered. To reflect the radioactive decay, four decay chains with the radionuclides included in the high-level radioactive wastes were selected. From the simulation results, the half-life and distribution coefficient in the rock matrix, as well as multiple pathways, had an influence on the mass flux of the radionuclides. For enhancing the reliability of safety assessment, this reveals that identifying the history of the radionuclides contained in the high-level wastes and investigating the sorption processes between the radionuclides and the rock matrix in the field condition are preferentially necessary.

Evaluation of nuclide release scenarios for a hypothetical LILW repository

A program for the safety assessment and performance evaluation of a low- and intermediate-level radioactive waste (LILW) repository system has been developed. Utilizing GoldSim (2006), the program evaluates nuclide release and transport into the geosphere and biosphere under various disruptive natural and manmade events and scenarios that can occur after a waste package failure. We envisaged and illustrated these events and scenarios as occurring after the closure of a hypothetical LILW repository, and they included the degradation of various manmade barriers, pumping well drilling, and natural disruptions such as the sudden formation of a preferential flow pathway in the far-field area of the repository. Possible enhancement of nuclide transport facilitated by colloids or chelating agents is also dealt with. We used the newly-developed GoldSim template program, which is capable of various nuclide release scenarios and is greatly suited for simulating a potential repository given the geological circumstances in Korea, to create the detailed source-term and near-field release scheme, various nuclide transport modes in the far-field geosphere area, and the biosphere transfer. Even though all parameter values applied to the hypothetical repository were assumed, the illustrative results, particularly the probabilistic calculations and sensitivity studies, may be informative under various scenarios. ( GoldSim, LILW, Nuclide transport, Safety assessment, Scenario).

Development of compartment models with Markov-chain processes for radionuclide transport in repository region

This paper presents a new radionuclide transport model for performance assessment and design of a geologic repository for high-level radioactive waste. The model uses compartmentalization of a model space and a Markov-chain process to describe the transport. The model space is divided into an array of compartments, among which a transition probability matrix describes radionuclide transport. While similar to the finite-difference method, it has several advantages such as flexibility to include various types of transport processes and reactions due to probabilistic interpretation, and higher-order accuracy resulting from direct formulation in a discrete-time frame. We demonstrated application of this model with a hypothetical repository in porous rock formation. First we calculated a three-dimensional steady-state heterogeneous groundwater flow field numerically by the finite-element method. The transition probability matrix was constructed based on the flow field and hydraulic dispersion coefficient. The present approach has been found to be effective in modeling radionuclide transport at a repository scale while taking into account the effects of change in hydraulic properties on the repository performance. Numerical exploration results indicate that engineered barrier configuration and material degradation have substantial effects on radionuclide release from the repository.

Geo-descriptive Modeling of Water Conducting Features Characterized in Sedimentary Formations in Horonobe Area of Japan

AbstractA three-dimensional (3-D) discrete fracture network (DFN) geo-descriptive model is developed for water conducting features (WCFs) in the sedimentary formations of Horonobe underground research laboratory (URL) in Japan. Fracturing and faulting system in/around the URL area, which is the main investigation area of the Horonobe URL project, is characterized by taking into account borehole geophysical logging data, regional geologic/structural data, and fracture/fault data (orientation, intensity, size) obtained from the surface-based investigations. Volumetric fracture intensity potential is estimated by the correlation and the multi-linear regression analysis of observed data, and is used as one of controls for 3-D DFN model. A regional scale 3-D geo-descriptive DFN model is constructed based on the analyzed fracturing system identified for the WCFs. The current 3-D geo-descriptive model could be utilized explicitly to derive performance assessment (PA) parameters for the hypothetical repository of the high-level radioactive wastes in Japan, and to assist optimization of the safe repository design.

A GoldSim model for the safety assessment of an HLW repository

An assessment program for the evaluation of a high-level waste (HLW) repository has been developed by utilizing GoldSim, by which nuclide transports in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release could be modeled and evaluated. To demonstrate its usability, three illustrative cases including the influence of a groundwater flow pattern through canisters associated with a flowing groundwater through fractures, and the possible disruptive events caused by an accidental human intrusion or an earthquake have been investigated and illustrated for a hypothetical Korean HLW repository. Even though all the parameter values applied to a hypothetical repository are assumed without any real base, the illustrative cases are very informative especially when seeing the result of the probabilistic calculation with the groundwater flow patterns which is turned out to be possibly important for nuclide release and further transport in and around the repository system.

A case study on the influence of THM coupling on the near field safety of a spent fuel repository in sparsely fractured granite

In order to demonstrate the feasibility of geological disposal of spent CANDU fuel in Canada, a safety assessment was performed for a hypothetical repository in the Canadian Shield. The assessment shows that the maximum long term radionuclide release from such repository would meet international criteria for dose rate; however, uncertainties in the assumed evolution of the repository were identified. Such uncertainties could be resolved by the consideration of coupled Thermal-Hydro-Mechanical-Chemical (THMC) processes. In Task A of the DECOVALEX-THMC project, THM models were developed within the framework of the theory of poroelasticity. Such model development was performed in an iterative manner, using experimental data from laboratory and field tests. The models were used to perform near-field simulations of the evolution of the repository in order to address the above-mentioned uncertainties. This paper presents the definition and rationale of task A and the results of the simulations. From a repository safety point of view, the simulations predict that the maximum temperature would be well below the design target of 100°C; however, the stress on the container can marginally exceed the design value of 15 MPa. However, the most important finding from the simulations is that a rock damage zone could form around the emplacement borehole. Such damage zone can extend a few metres from the walls of the emplacement holes, with permeability values that are orders of magnitude higher than the initial values. The damage zone has the potential to increase the radionuclide transport flux from the geosphere; the effect of such an increase should be taken into account in the safety assessment and mitigated if necessary by the provision of sealing systems.

Performance Assessment of Low-Level Waste Disposal Facilities Using Coupled Unsaturated Flow and Reactive Transport Simulators

ABSTRACTRecent advances in the development of reactive chemical transport simulators have made it possible to use these tools in performance assessments (PAs) for nuclear waste disposal. Reactive transport codes were used to evaluate the impacts of design modifications on the performance of two shallow subsurface disposal systems for low-level radioactive waste. The first disposal system, located at the Hanford site in Richland, Washington, is for disposal of lowlevel waste glass. Glass waste blocks will be disposed in subsurface trenches, surrounded by backfill material. Using different waste package sizes and layering had a small impact on technetium release rates to the vadose zone. The second disposal system involves a hypothetical repository for low-level waste in Italy. A model of uranium release from a grout waste form was developed using the STORM reactive transport code. Uranium is predicted to be relatively insoluble for several hundred years under the high-pH environment of the cement pore water. The effect of using different filler materials between the waste packages on uranium flux to the vadose zone proved to have a negligible impact on release rates.

Comparison of alternative modelling approaches for groundwater flow in fractured rock

In performance assessment studies of radioactive waste disposal in crystalline rocks, one source of uncertainty is the appropriateness of conceptual models of the physical processes contributing to the potential transport of radionuclides. The Alternative Models Project (AMP) evaluates the uncertainty of models of groundwater flow, an uncertainty that arises from alternative conceptualisations of groundwater movement in fractured media. The AMP considers three modelling approaches for simulating flow and advective transport from the waste canisters to the biosphere: Stochastic Continuum, Discrete Fracture Network, and Channel Network. Each approach addresses spatial variability via Monte Carlo simulation, whose realisations are summarised by the statistics of three simplified measures of geosphere performance: travel time, transport resistance (a function of travel distance, flow-wetted surface per volume of rock, and Darcy velocity along a flowpath), and canister flux (Darcy velocity at repository depth). The AMP uses a common reference case defined by a specific model domain, boundary conditions, and layout of a hypothetical repository, with a consistent set of summary statistics to facilitate the comparison of the three approaches. The three modelling approaches predict similar median travel times and median canister fluxes, but dissimilar variability. The three modelling approaches also predict similar values for minimum travel time and maximum canister flux, and predict similar locations for particles exiting the geosphere. The results suggest that the problem specifications (i.e. boundary conditions and gross hydrogeology) constrain the flow modelling, limiting the impact of this conceptual uncertainty on performance assessment.

Radionuclide Transport Calculations in the Safety Assessment SR 97

ABSTRACTThis study treats radionuclide transport calculations for a canister defect scenario in the safety assessment SR 97, which concerns a deep repository for spent nuclear fuel of the KBS-3 type in Sweden. The aims of the calculations are to:Quantitatively describe the radionuclide transport.Show the impact of uncertainty in input data and show which parameters govern the calculated release rates.Compare three different real sites in Sweden (Aberg, Beberg and Ceberg) with each other and with dose limits given in Swedish regulations (none of the sites is considered in the on-going localization process). Only briefly described in this paper.Illustrate the impact of the different barriers in the system.Deterministic calculations illustrate the radionuclide transport for reasonable conditions. Uncertainty cases show the influence of the uncertainty for data related to different parts of the repository system by systematically giving them pessimistic values while all others are reasonable. Simplified probabilistic calculations have also been performed.The analysis shows that the most important parameters in the near field are the number of defective canisters and the instant release fraction. In the far field the most important uncertainties affecting release and retention are connected to permeability and connectivity of the fractures in the rock. The dose rate in the biosphere is essentially controlled by the possibilities of dilution.The calculated maximum doses for the hypothetical repositories are well below the dose limits, and hence they meet the acceptance criteria for a deep repository for spent fuel.

A Stochastic‐Analytical Framework for Safety Assessment of Waste Repositories: 2. Application

AbstractAn analytical framework for probabilistic safety assessment of geological repositories is applied for a high‐level nuclear waste repository. The framework is based on stochastic‐analytical solution techniques and offers a versatile tool for the analysis of various scenarios. Features, events, and processes for a hypothetical repository, geosphere, and biosphere are considered. Explicit solutions in terms of the expected value and variance of the mass flux or cumulative mass flux are derived for each case considered. The uncertainty pertaining to features and events results in increased time spans of elevated hazard levels with possible exceedance of given regulatory criteria. Mass transfer processes, such as sorption and matrix diffusion, may result in significant reductions of radioactivity fluxes and subsequent doses in the biosphere.

Accommodating ground water velocity uncertainties in the advection-dispersion approach to geologic nuclear waste migration

This note shows how uncertainties in nearfield and farfield ground water velocities affect the inventory that migrates from a geologic nuclear waste repository within the classical advection-dispersion approach and manifest themselves through both the finite variances and covariances in the activities of transported nuclides and in the apparent scale dependence of the host rock's dispersivity. Included is a demonstration of these effects for an actinide chain released from used CANDU™ fuel buried in a hypothetical repository.

The transport of radionuclides discharged from a geologic waste repository

A 3-D conservative solution to the convection-dispersion-retardation equation describing the migration of radionuclides from a geologic nuclear waste disposal site is derived using a Green's function approach under conditions in which the contaminant inventory is preferentially discharged to the steady groundwater flow. The numerical implementation of the upper estimate to the transported inventory, as embodied in the CANUKE© code, is described; its applicability is demonstrated by calculating the groundwater concentration of the neptunium series nuclides released from used CANDU™ fuel emplaced in an hypothetical repository; and its general usage is discussed. The ability to handle actinide chains of arbitrary length, whatever their initial inventories, yields both less conservative and physically more realistic estimates of transported inventories than schemes which begin with a contracted chain of longer-lived nuclides whose progeny are assumed to accompany them in their migration, irrespective of their relative mobilities.

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 > 10 5 kg/cm 2 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 2 1 2 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.