Publications

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Deep Borehole Disposal (DBD) of radioactive waste has some clear advantages over mined repositories, including incremental construction and loading, enhanced natural barriers provided by deep continental crystalline basement, and reduced site characterization. Unfavorable features for a DBD site include upward vertical fluid potential gradients, presence of economically exploitable natural resources, presence of high permeability connection from the waste disposal zone to the shallow subsurface, and significant probability of future volcanic activity. Site characterization activities would encompass geomechanical (i.e., rock stress state, fluid pressure, and faulting), geological (i.e., both overburden and bedrock lithology), hydrological (i.e., quantity of fluid, fluid convection properties, and solute transport mechanisms), chemical (i.e., rock and fluid interaction), and socioeconomic (i.e., likelihood for human intrusion) aspects. For a planned Deep Borehole Field Test (DBFT), site features and/or physical processes would be evaluated using both direct (i.e., sampling and in-hole testing) and indirect (i.e., surface and borehole geophysical) methods for efficient and effective characterization. Surface-based characterization would be used to guide the exploratory drilling program, once a candidate DBFT site has been selected. Borehole based characterization will be used to determine the variability of system state (i.e., stress, pressure, temperature, petrology, and water chemistry) with depth, and to develop material and system parameters relevant for numerical simulation. While the site design of DBD could involve an array of disposal boreholes, it may not be necessary to characterize each borehole in detail. Characterization strategies will be developed in the DBFT that establish disposal system safety sufficient for licensing a disposal array.

While deep borehole disposal of nuclear waste should rely primarily on off-the-shelf technologies pioneered by the oil and gas and geothermal industries, the development of new science and technology will remain important. Key knowledge gaps have been outlined in the research roadmap for deep boreholes (B. Arnold et al, 2012, Research, Development, and Demonstration Roadmap for Deep Borehole Disposal, Sandia National Laboratories, SAND2012-8527P) and in a recent Deep Borehole Science Needs Workshop. Characterizing deep crystalline basement, understanding the nature and role of deep fractures, more precisely age-dating deep groundwaters, and demonstrating long-term performance of seals are all important topics of interest. Overlapping deep borehole and enhanced geothermal technology needs include: quantification of seal material performance/failure, stress measurement beyond the borehole, advanced drilling and completion tools, and better subsurface sensors. A deep borehole demonstration has the potential to trigger more focused study of deep hydrology, high temperature brine-rock interaction, and thermomechanical behavior.

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Report RWEV-REP-001, Analysis of Postclosure Groundwater Impacts for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High Level Radioactive Waste at Yucca Mountain, Nye County, Nevada was issued by the DOE in 2009 and is currently being updated. Sandia National Laboratories (SNL) provided support for the original document, performing calculations and extracting data from the Yucca Mountain Performance Assessment Model that were used as inputs to the contaminant transport and dose calculations by Jason Associates Corporation, the primary developers of the DOE report. The inputs from SNL were documented in LSA-AR-037, Inputs to Jason Associates Corporation in Support of the Postclosure Repository Supplemental Environmental Impact Statement. To support the updating of the original Groundwater Impacts document, SNL has reviewed the inputs provided in LSA-AR-037 to verify that they are current and appropriate for use. The results of that assessment are documented here.

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In this study, we use PFLOTRAN, a highly scalable, parallel, flow and reactive transport code to simulate the concentrations of 3H, 3He, CFC-11, CFC-12, CFC-113, SF6, 39Ar, 81Kr, 4He and themean groundwater age in heterogeneous fields on grids with an excess of 10 million nodes. We utilize this computational platform to simulate the concentration of multiple tracers in high-resolution, heterogeneous 2-D and 3-D domains, and calculate tracer-derived ages. Tracer-derived ages show systematic biases toward younger ages when the groundwater age distribution contains water older than the maximum tracer age. The deviation of the tracer-derived age distribution from the true groundwater age distribution increases with increasing heterogeneity of the system. However, the effect of heterogeneity is diminished as the mean travel time gets closer the tracer age limit. Age distributions in 3-D domains differ significantly from 2-D domains. 3D simulations show decreased mean age, and less variance in age distribution for identical heterogeneity statistics. High-performance computing allows for investigation of tracer and groundwater age systematics in high-resolution domains, providing a platform for understanding and utilizing environmental tracer and groundwater age information in heterogeneous 3-D systems. Groundwater environmental tracers can provide important constraints for the calibration of groundwater flow models. Direct simulation of environmental tracer concentrations in models has the additional advantage of avoiding assumptions associated with using calculated groundwater age values. This study quantifies model uncertainty reduction resulting from the addition of environmental tracer concentration data. The analysis uses a synthetic heterogeneous aquifer and the calibration of a flow and transport model using the pilot point method. Results indicate a significant reduction in the uncertainty in permeability with the addition of environmental tracer data, relative to the use of hydraulic measurements alone. Anthropogenic tracers and their decay products, such as CFC11, 3H, and 3He, provide significant constraint oninput permeability values in the model. Tracer data for 39Ar provide even more complete information on the heterogeneity of permeability and variability in the flow system than the anthropogenic tracers, leading to greater parameter uncertainty reduction.

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Deep boreholes could be a relatively inexpensive, safe, and rapidly deployable strategy for disposing Americas nuclear waste. To study this approach, Sandia invested in a three year LDRD project entitled %E2%80%9CRadionuclide Transport from Deep Boreholes.%E2%80%9D In the first two years, the borehole reference design and backfill analysis were completed and the supporting modeling of borehole temperature and fluid transport profiles were done. In the third year, some of the logistics of implementing a deep borehole waste disposal system were considered. This report describes what was learned in the third year of the study and draws some conclusions about the potential bottlenecks of system implementation.

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A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall, the results of the reference design development and the cost analysis support the technical feasibility of the deep borehole disposal concept for high-level radioactive waste.

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Disposal of high-level radioactive waste, including spent nuclear fuel, in deep (3 to 5 km) boreholes is a potential option for safely isolating these wastes from the surface and near-surface environment. Existing drilling technology permits reliable and cost-effective construction of such deep boreholes. Conditions favorable for deep borehole disposal in crystalline basement rocks, including low permeability, high salinity, and geochemically reducing conditions, exist at depth in many locations, particularly in geologically stable continental regions. Isolation of waste depends, in part, on the effectiveness of borehole seals and potential alteration of permeability in the disturbed host rock surrounding the borehole. Coupled thermal-mechanical-hydrologic processes induced by heat from the radioactive waste may impact the disturbed zone near the borehole and borehole wall stability. Numerical simulations of the coupled thermal-mechanical response in the host rock surrounding the borehole were conducted with three software codes or combinations of software codes. Software codes used in the simulations were FEHM, JAS3D, Aria, and Adagio. Simulations were conducted for disposal of spent nuclear fuel assemblies and for the higher heat output of vitrified waste from the reprocessing of fuel. Simulations were also conducted for both isotropic and anisotropic ambient horizontal stress in the host rock. Physical, thermal, and mechanical properties representative of granite host rock at a depth of 4 km were used in the models. Simulation results indicate peak temperature increases at the borehole wall of about 30 C and 180 C for disposal of fuel assemblies and vitrified waste, respectively. Peak temperatures near the borehole occur within about 10 years and decline rapidly within a few hundred years and with distance. The host rock near the borehole is placed under additional compression. Peak mechanical stress is increased by about 15 MPa (above the assumed ambient isotropic stress of 100 MPa) at the borehole wall for the disposal of fuel assemblies and by about 90 MPa for vitrified waste. Simulated peak volumetric strain at the borehole wall is about 420 and 2600 microstrain for the disposal of fuel assemblies and vitrified waste, respectively. Stress and volumetric strain decline rapidly with distance from the borehole and with time. Simulated peak stress at and parallel to the borehole wall for the disposal of vitrified waste with anisotropic ambient horizontal stress is about 440 MPa, which likely exceeds the compressive strength of granite if unconfined by fluid pressure within the borehole. The relatively small simulated displacements and volumetric strain near the borehole suggest that software codes using a nondeforming grid provide an adequate approximation of mechanical deformation in the coupled thermal-mechanical model. Additional modeling is planned to incorporate the effects of hydrologic processes coupled to thermal transport and mechanical deformation in the host rock near the heated borehole.

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Preliminary evaluation of deep borehole disposal of high-level radioactive waste and spent nuclear fuel indicates the potential for excellent long-term safety performance at costs competitive with mined repositories. Significant fluid flow through basement rock is prevented, in part, by low permeabilities, poorly connected transport pathways, and overburden self-sealing. Deep fluids also resist vertical movement because they are density stratified. Thermal hydrologic calculations estimate the thermal pulse from emplaced waste to be small (less than 20 C at 10 meters from the borehole, for less than a few hundred years), and to result in maximum total vertical fluid movement of {approx}100 m. Reducing conditions will sharply limit solubilities of most dose-critical radionuclides at depth, and high ionic strengths of deep fluids will prevent colloidal transport. For the bounding analysis of this report, waste is envisioned to be emplaced as fuel assemblies stacked inside drill casing that are lowered, and emplaced using off-the-shelf oilfield and geothermal drilling techniques, into the lower 1-2 km portion of a vertical borehole {approx}45 cm in diameter and 3-5 km deep, followed by borehole sealing. Deep borehole disposal of radioactive waste in the United States would require modifications to the Nuclear Waste Policy Act and to applicable regulatory standards for long-term performance set by the US Environmental Protection Agency (40 CFR part 191) and US Nuclear Regulatory Commission (10 CFR part 60). The performance analysis described here is based on the assumption that long-term standards for deep borehole disposal would be identical in the key regards to those prescribed for existing repositories (40 CFR part 197 and 10 CFR part 63).

This paper summarizes the numerical site scale model developed to simulate the transport of radionuclides via ground water in the saturated zone beneath Yucca Mountain.

Simulation of potential radionuclide transport in the saturated zone from beneath the proposed repository at Yucca Mountain to the accessible environment is an important aspect of the total system performance assessment (TSPA) for disposal of high-level radioactive waste at the site. Analyses of uncertainty and sensitivity are integral components of the TSPA and have been conducted at both the sub-system and system levels to identify parameters and processes that contribute to the overall uncertainty in predictions of repository performance. Results of the sensitivity analyses indicate that uncertainty in groundwater specific discharge along the flow path in the saturated zone from beneath the repository is an important contributor to uncertainty in TSPA results and is the dominant source of uncertainty in transport times in the saturated zone for most radionuclides. Uncertainties in parameters related to matrix diffusion in the volcanic units, colloid-facilitated transport, and sorption are also important contributors to uncertainty in transport times to differing degrees for various radionuclides.

Sandia National Laboratories, New Mexico (SNL/NM) is a government-owned/contractor-operated facility. Sandia Corporation (Sandia), a wholly owned subsidiary of Lockheed Martin Corporation, manages and operates the laboratory for the U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA). The DOE/NNSA Sandia Site Office (SSO) administers the contract and oversees contractor operations at the site. This annual report summarizes data and the compliance status of Sandia Corporation’s environmental protection and monitoring programs through December 31, 2007. Major environmental programs include air quality, water quality, groundwater protection, terrestrial surveillance, waste management, pollution prevention (P2), environmental restoration (ER), oil and chemical spill prevention, and implementation of the National Environmental Policy Act (NEPA). Environmental monitoring and surveillance programs are required by DOE Order 450.1, Environmental Protection Program (DOE 2007a) and DOE Manual 231.1-1A, Environment, Safety, and Health Reporting (DOE 2007).

Uncertainty in site characterization arises from a lack of data and knowledge about a site and includes uncertainty in the boundary conditions, uncertainty in the characteristics, location, and behavior of major features within an investigation area (e.g., major faults as barriers or conduits), uncertainty in the geologic structure, as well as differences in numerical implementation (e.g., 2-D versus 3-D, finite difference versus finite element, grid resolution, deterministic versus stochastic, etc.). Since the true condition at a site can never be known, selection of the best conceptual model is very difficult. In addition, limiting the understanding to a single conceptualization too early in the process, or before data can support that conceptualization, may lead to confidence in a characterization that is unwarranted as well as to data collection efforts and field investigations that are misdirected and/or redundant. Using a series of numerical modeling experiments, this project examined the application and use of information criteria within the site characterization process. The numerical experiments are based on models of varying complexity that were developed to represent one of two synthetically developed groundwater sites; (1) a fully hypothetical site that represented a complex, multi-layer, multi-faulted site, and (2) a site that was based on the Horonobe site in northern Japan. Each of the synthetic sites were modeled in detail to provide increasingly informative 'field' data over successive iterations to the representing numerical models. The representing numerical models were calibrated to the synthetic site data and then ranked and compared using several different information criteria approaches. Results show, that for the early phases of site characterization, low-parameterized models ranked highest while more complex models generally ranked lowest. In addition, predictive capabilities were also better with the low-parameterized models. For the latter iterations, when more data were available, the information criteria rankings tended to converge on the higher parameterized models. Analysis of the numerical experiments suggest that information criteria rankings can be extremely useful for site characterization, but only when the rankings are placed in context and when the contribution of each bias term is understood.

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Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in the assessment of radioactive waste disposal and at the time of this publication is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. In countries with small radioactive waste programs, international technology transfer program efforts are often hampered by small budgets, schedule constraints, and a lack of experienced personnel. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available software codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission (NRC) and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, revitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a credible and solid computational platform for constructing probabilistic safety assessment models. This document is a reference users guide for the GoldSim/BLT-MS integrated modeling software package developed as part of a cooperative technology transfer project between Sandia National Laboratories and the Institute of Nuclear Energy Research (INER) in Taiwan for the preliminary assessment of several candidate low-level waste repository sites. Breach, Leach, and Transport-Multiple Species (BLT-MS) is a U.S. NRC sponsored code which simulates release and transport of contaminants from a subsurface low-level waste disposal facility. GoldSim is commercially available probabilistic software package that has radionuclide transport capabilities. The following report guides a user through the steps necessary to use the integrated model and presents a successful application of the paradigm of renewing legacy codes for contemporary application.

Sandia National Laboratories and the Institute of Nuclear Energy Research, Taiwan have collaborated in a technology transfer program related to low-level radioactive waste (LLW) disposal in Taiwan. Phase I of this program included regulatory analysis of LLW final disposal, development of LLW disposal performance assessment capabilities, and preliminary performance assessments of two potential disposal sites. Performance objectives were based on regulations in Taiwan and comparisons to those in the United States. Probabilistic performance assessment models were constructed based on limited site data using software including GoldSim, BLT-MS, FEHM, and HELP. These software codes provided the probabilistic framework, container degradation, waste-form leaching, groundwater flow, radionuclide transport, and cover infiltration simulation capabilities in the performance assessment. Preliminary performance assessment analyses were conducted for a near-surface disposal system and a mined cavern disposal system at two representative sites in Taiwan. Results of example calculations indicate peak simulated concentrations to a receptor within a few hundred years of LLW disposal, primarily from highly soluble, non-sorbing radionuclides.

This report represents the final product of a background literature review conducted for the Nuclear Waste Management Organization of Japan (NUMO) by Sandia National Laboratories, Albuquerque, New Mexico, USA. Internationally, research of hydrological and transport processes in the context of high level waste (HLW) repository performance, has been extensive. However, most of these studies have been conducted for sites that are within tectonically stable regions. Therefore, in support of NUMO's goal of selecting a site for a HLW repository, this literature review has been conducted to assess the applicability of the output from some of these studies to the geological environment in Japan. Specifically, this review consists of two main tasks. The first was to review the major documents of the main HLW repository programs around the world to identify the most important hydrologic and transport parameters and processes relevant in each of these programs. The review was to assess the relative importance of processes and measured parameters to site characterization by interpretation of existing sensitivity analyses and expert judgment in these documents. The second task was to convene a workshop to discuss the findings of Task 1 and to prioritize hydrologic and transport parameters in the context of the geology of Japan. This report details the results and conclusions of both of these Tasks.

Sandia National Laboratories (SNL) and Taiwan's Institute for Nuclear Energy Research (INER) have teamed together to evaluate several candidate sites for Low-Level Radioactive Waste (LLW) disposal in Taiwan. Taiwan currently has three nuclear power plants, with another under construction. Taiwan also has a research reactor, as well as medical and industrial wastes to contend with. Eventually the reactors will be decomissioned. Operational and decommissioning wastes will need to be disposed in a licensed disposal facility starting in 2014. Taiwan has adopted regulations similar to the US Nuclear Regulatory Commission's (NRC's) low-level radioactive waste rules (10 CFR 61) to govern the disposal of LLW. Taiwan has proposed several potential sites for the final disposal of LLW that is now in temporary storage on Lanyu Island and on-site at operating nuclear power plants, and for waste generated in the future through 2045. The planned final disposal facility will have a capacity of approximately 966,000 55-gallon drums. Taiwan is in the process of evaluating the best candidate site to pursue for licensing. Among these proposed sites there are basically two disposal concepts: shallow land burial and cavern disposal. A representative potential site for shallow land burial is located on a small island in the Taiwan Strait with basalt bedrock and interbedded sedimentary rocks. An engineered cover system would be constructed to limit infiltration for shallow land burial. A representative potential site for cavern disposal is located along the southeastern coast of Taiwan in a tunnel system that would be about 500 to 800 m below the surface. Bedrock at this site consists of argillite and meta-sedimentary rocks. Performance assessment analyses will be performed to evaluate future performance of the facility and the potential dose/risk to exposed populations. Preliminary performance assessment analyses will be used in the site-selection process and to aid in design of the disposal system. Final performance assessment analyses will be used in the regulatory process of licensing a site. The SNL/INER team has developed a performance assessment methodology that is used to simulate processes associated with the potential release of radionuclides to evaluate these sites. The following software codes are utilized in the performance assessment methodology: GoldSim (to implement a probabilistic analysis that will explicitly address uncertainties); the NRC's Breach, Leach, and Transport - Multiple Species (BLT-MS) code (to simulate waste-container degradation, waste-form leaching, and transport through the host rock); the Finite Element Heat and Mass Transfer code (FEHM) (to simulate groundwater flow and estimate flow velocities); the Hydrologic Evaluation of Landfill performance Model (HELP) code (to evaluate infiltration through the disposal cover); the AMBER code (to evaluate human health exposures); and the NRC's Disposal Unit Source Term -- Multiple Species (DUST-MS) code (to screen applicable radionuclides). Preliminary results of the evaluations of the two disposal concept sites are presented.

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Natural gas is a clean fuel that will be the most important domestic energy resource for the first half the 21st centtuy. Ensuring a stable supply is essential for our national energy security. The research we have undertaken will maximize the extractable volume of gas while minimizing the environmental impact of surface disturbances associated with drilling and production. This report describes a methodology for comprehensive evaluation and modeling of the total gas system within a basin focusing on problematic horizontal fluid flow variability. This has been accomplished through extensive use of geophysical, core (rock sample) and outcrop data to interpret and predict directional flow and production trends. Side benefits include reduced environmental impact of drilling due to reduced number of required wells for resource extraction. These results have been accomplished through a cooperative and integrated systems approach involving industry, government, academia and a multi-organizational team within Sandia National Laboratories. Industry has provided essential in-kind support to this project in the forms of extensive core data, production data, maps, seismic data, production analyses, engineering studies, plus equipment and staff for obtaining geophysical data. This approach provides innovative ideas and technologies to bring new resources to market and to reduce the overall environmental impact of drilling. More importantly, the products of this research are not be location specific but can be extended to other areas of gas production throughout the Rocky Mountain area. Thus this project is designed to solve problems associated with natural gas production at developing sites, or at old sites under redevelopment.

Preliminary investigation areas (PIA) for a potential repository of high-level radioactive waste must be evaluated by NUMO with regard to a number of qualifying factors. One of these factors is related to earthquakes and fault activity. This study develops a spatial statistical assessment method that can be applied to the active faults in Japan to perform such screening evaluations. This analysis uses the distribution of seismicity near faults to define the width of the associated process zone. This concept is based on previous observations of aftershock earthquakes clustered near active faults and on the assumption that such seismic activity is indicative of fracturing and associated impacts on bedrock integrity. Preliminary analyses of aggregate data for all of Japan confirmed that the frequency of earthquakes is higher near active faults. Data used in the analysis were obtained from NUMO and consist of three primary sources: (1) active fault attributes compiled in a spreadsheet, (2) earthquake hypocenter data, and (3) active fault locations. Examination of these data revealed several limitations with regard to the ability to associate fault attributes from the spreadsheet to locations of individual fault trace segments. In particular, there was no direct link between attributes of the active faults in the spreadsheet and the active fault locations in the GIS database. In addition, the hypocenter location resolution in the pre-1983 data was less accurate than for later data. These pre-1983 hypocenters were eliminated from further analysis.

A probabilistic model and software tool has been developed to assist in the selection, design, and monitoring of long-term covers for contaminated subsurface sites. The software platform contains multiple modules that can be used to simulate relevant features, events, and processes that include water flux through the cover, source-term release, vadose-zone transport, saturated-zone transport, gas transport, and exposure pathways. The component models are integrated into a probabilistic total-system performance-assessment model within a drag-and-drop software platform. Uncertainty and sensitivity analyses can be conducted that yield the following primary benefits: (1) quantification of uncertainty in the simulated performance metrics; (2) identification of parameters most important to performance; and (3) comparison of alternative designs to optimize cost and performance. A case study has been performed using the Monticello Mill Tailings Site in Utah to illustrate the important features and benefits of the modeling approach and software. © 2003 Elsevier Ltd. All rights reserved.

A probabilistic, risk-based performance-assessment methodology has been developed to assist designers, regulators, and stakeholders in the selection, design, and monitoring of long-term covers for contaminated subsurface sites. This report describes the method, the software tools that were developed, and an example that illustrates the probabilistic performance-assessment method using a repository site in Monticello, Utah. At the Monticello site, a long-term cover system is being used to isolate long-lived uranium mill tailings from the biosphere. Computer models were developed to simulate relevant features, events, and processes that include water flux through the cover, source-term release, vadose-zone transport, saturated-zone transport, gas transport, and exposure pathways. The component models were then integrated into a total-system performance-assessment model, and uncertainty distributions of important input parameters were constructed and sampled in a stochastic Monte Carlo analysis. Multiple realizations were simulated using the integrated model to produce cumulative distribution functions of the performance metrics, which were used to assess cover performance for both present- and long-term future conditions. Performance metrics for this study included the water percolation reaching the uranium mill tailings, radon gas flux at the surface, groundwater concentrations, and dose. Results from uncertainty analyses, sensitivity analyses, and alternative design comparisons are presented for each of the performance metrics. The benefits from this methodology include a quantification of uncertainty, the identification of parameters most important to performance (to prioritize site characterization and monitoring activities), and the ability to compare alternative designs using probabilistic evaluations of performance (for cost savings).

A probabilistic, risk-based performance-assessment methodology is being developed to assist designers, regulators, and involved stakeholders in the selection, design, and monitoring of long-term covers for contaminated subsurface sites. This report presents an example of the risk-based performance-assessment method using a repository site in Monticello, Utah. At the Monticello site, a long-term cover system is being used to isolate long-lived uranium mill tailings from the biosphere. Computer models were developed to simulate relevant features, events, and processes that include water flux through the cover, source-term release, vadose-zone transport, saturated-zone transport, gas transport, and exposure pathways. The component models were then integrated into a total-system performance-assessment model, and uncertainty distributions of important input parameters were constructed and sampled in a stochastic Monte Carlo analysis. Multiple realizations were simulated using the integrated model to produce cumulative distribution functions of the performance metrics, which were used to assess cover performance for both present- and long-term future conditions. Performance metrics for this study included the water percolation reaching the uranium mill tailings, radon flux at the surface, groundwater concentrations, and dose. Results of this study can be used to identify engineering and environmental parameters (e.g., liner properties, long-term precipitation, distribution coefficients) that require additional data to reduce uncertainty in the calculations and improve confidence in the model predictions. These results can also be used to evaluate alternative engineering designs and to identify parameters most important to long-term performance.

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The effective-porosity approach and the dual-porosity approach are examined as two alternative conceptual models of radionuclide migration in fractured media of the saturated zone at Yucca Mountain. Numerical simulations of onedimensional radionuclide transport are performed for the domain relevant to repository performance assessment using the two alternative conceptual approaches. Dual-porosity solute transport modeling produces similar results to effective-porosity modeling for fracture spacing of less than approximately 1 m and greater than about 200 m, which corresponds to values of effective porosity equal to the matrix porosity and the fracture porosity, respectively. For intermediate values of fracture spacing, the dual-porosity approach results in concentration breakthrough curves that differ significantly from the effectiveporosity approach and are characterized by earlier first arrival, greater apparent dispersion, and lower concentrations at later times. The effective-porosity approach, as implemented in recent performance assessment analyses of saturated zone transport at Yucca Mountain, is conservative compared to the dual-porosity approach in terms of both radionuclide concentrations and, generally, travel times.