ANS IHLRWM 2017 - 16th International High-Level Radioactive Waste Management Conference: Creating a Safe and Secure Energy Future for Generations to Come - Driving Toward Long-Term Storage and Disposal
Transportation of spent nuclear fuel (SNF) is expected to increase in the future, as the nuclear fuel infrastructure continues to expand and fuel takeback programs increase in popularity. Analysis of potential risks and threats to SNF shipments is currently performed separately for safety and security. However, as SNF transportation increases, the plausible threats beyond individual categories and the interactions between them become more apparent. A new approach is being developed to integrate safety, security, and safeguards (3S) under a system-theoretic framework and a probabilistic risk framework. At the first stage, a simplified scenario will be implemented using a dynamic probabilistic risk assessment (DPRA) method. This scenario considers a rail derailment followed by an attack. The consequences of derailment are calculated with RADTRAN, a transportation risk analysis code. The attack scenarios are analyzed with STAGE, a combat simulation model. The consequences of the attack are then calculated with RADTRAN. Note that both accident and attack result in SNF cask damage and a potential release of some fraction of the SNF inventory into the environment. The major purpose of this analysis was to develop the input data for DPRA. Generic PWR and BWR transportation casks were considered. These data were then used to demonstrate the consequences of hypothetical accidents in which the radioactive materials were released into the environment. The SNF inventory is one of the most important inputs into the analysis. Several pressurized water reactor (PWR) and boiling water reactor (BWR) fuel burnups and discharge times were considered for this proof-of-concept. The inventory was calculated using ORIGEN (point depletion and decay computer code, Oak Ridge National Laboratory) for 3 characteristic burnup values (40, 50, and 60 GWD/MTU) and 4 fuel ages (5, 10, 25 and 50 years after discharge). The major consequences unique to the transportation of SNF for both accident and attack are the results of the dispersion of radionuclides in the environment. The dynamic atmospheric dispersion model in RADTRAN was used to calculate these consequences. The examples of maximum exposed individual (MEI) dose, early mortality and soil contamination are discussed to demonstrate the importance of different factors. At the next stage, the RADTRAN outputs will be converted into a form compatible with the STAGE analysis. As a result, identification of additional risks related to the interaction between characteristics becomes a more straightforward task. In order to present the results of RADTRAN analysis in a framework compatible with the results of the STAGE analysis, the results will be grouped into three categories: • Immediate negative harms •Future benefits that cannot be realized •Additional increases in future risk By describing results within generically applicable categories, the results of safety analysis are able to be placed in context with the risk arising from security events.
ANS IHLRWM 2017 - 16th International High-Level Radioactive Waste Management Conference: Creating a Safe and Secure Energy Future for Generations to Come - Driving Toward Long-Term Storage and Disposal
Disposal of high-level radioactive waste in a deep geological repository in crystalline host rock is one of the potential options for long term isolation. Characterization of the natural barrier system is an important component of the disposal option. In this study we present numerical modeling of flow and transport in fractured crystalline rock using an updated fracture continuum model (FCM). The FCM is a stochastic method that maps the permeability of discrete fractures onto a regular grid. The original method [1] has been updated to provide capabilities that enhance representation of fractured rock. A companion paper [2] provides details of the methods for generating fracture network. In this paper use of the fracture model for the simulation of flow and transport is presented. Simulations were conducted to estimate flow and transport using an enhanced FCM method. Distributions of fracture parameters were used to generate a selected number of realizations. For each realization FCM produced permeability and porosity fields. The PFLOTRAN code [3] was used to simulate flow and transport. Simulation results and analysis are presented. The results indicate that the FCM approach is a viable method to model fractured crystalline rocks. The FCM is a computationally efficient way to generate realistic representation of complex fracture systems. This approach is of interest to nuclear waste disposal modeling applied over large domains.
ANS IHLRWM 2017 - 16th International High-Level Radioactive Waste Management Conference: Creating a Safe and Secure Energy Future for Generations to Come - Driving Toward Long-Term Storage and Disposal
The Deep Borehole Field Test will include demonstration of the emplacement and retrieval of test waste packages (containing no waste) in a 5 km deep borehole drilled into the crystalline basement. A conceptual design for packaging, surface handling and transfer equipment, and borehole emplacement was developed in anticipation of the demonstration project. Test packages are designed to withstand external pressure greater than 65 MPa, at temperature up to 170°C. Two packaging concepts were developed: 1) flasktype for granular waste, and 2) internal semi-flush type for waste that is pre-canistered in cylindrical containers. Oilfield casing materials and sealing connections would be selected giving a safety factor of 2.0 against yield. Packages would have threaded fittings top and bottom for attachment of impact limiters and latch fittings. Packages would be lowered one-at-a-time into the borehole on electric wireline. This offers important safety advantages over using drill pipe or coiled tubing to lower waste packages, because it avoids the possibility of dropping a heavy assembly in the borehole. An electromechanical latch would release each package, or reconnect for retrieval. Frequency of waste package delivery to a disposal site could be the effective limit on emplacement throughput. Packages would be delivered in a shielded Type B transportation cask and transferred to a shielded, doubleended transfer cask on site. The transfer cask would be upended over the borehole and secured to the wellhead. The transfer cask would become an integral part of the pressure control envelope for well pressure control. Blowout preventers can be incorporated as needed for regulatory compliance. Operational safety has been assessed with respect to normal operations, and off-normal events that could cause package breach in the borehole. Worker exposures can be limited by using standard industry practices for nuclear material handling. The waste packages would effectively be robust pressure vessels that will not breach if dropped during surface handling. The possibility of package breach in the borehole during emplacement can be effectively eliminated using impact limiters on every package.
ANS IHLRWM 2017 - 16th International High-Level Radioactive Waste Management Conference: Creating a Safe and Secure Energy Future for Generations to Come - Driving Toward Long-Term Storage and Disposal
Understanding subsurface fracture network properties at the field scale is important for a number of environmental and economic problems, including siting of spent nuclear fuel repositories, geothermal exploration, and many others. This typically encompasses large volumes of fractured rocks with the properties inferred from the observations at rock outcrops and, if available, from the measurements in exploratory boreholes, quarries, and tunnels. These data are inherently spatially limited and a stochastic model is required to extrapolate the fracture properties over the large volumes of rocks. This study (1) describes three different methods of generating fracture networks developed for use in the fractured continuum model (FCM) and (2) provides a few examples of how these methods impact the predictions of simulated groundwater transport. A detailed analysis of the transport simulations using FCM is provided in the separate paper by the same authors (to be presented at IHLRWM 2017 conference). FCM is based on the effective continuum approaches modified to represent fractures. The permeability of discrete fractures is mapped onto a regular three-dimensional grid. The x-, y-, and z effective permeability values of a grid block are calculated from the tensor. The tensor parameters are fracture aperture, dip, strike, and number of fractures in the grid block (spacing). All three methods use the fracture properties listed above to generate corresponding permeability fields. However, the assumptions and conceptual representation of fracture network from which these properties are derived are very different. The Sequential Gaussian Simulation (SGSim) method does not require an assumption regarding the fracture shape. Fracture aperture, spacing, and orientation are defined based on the field observations. Spatially correlated features (continuation of fracture in the direction of the orientation) are created using spatially correlated random numbers generated with SGSIM code. With this method an exact number of fractures cannot be generated. The Ellipsim method assumes that the fractures are two-dimensional elliptical shapes that can be described with radius and aspect ratio. The knowledge of the fracture (ellipse) radius probability distribution is required. The fracture aperture is calculated from the ellipse radius. For this option an exact number of fractures can be generated. The fracture networks generated with SGSim and Ellipsim are not necessarily connected. The connectivity is achieved indirectly via matrix permeability that can be viewed as the permeability of much smaller fractions. The discrete fracture network (DFN) generator assumes elliptical fracture shapes and requires the same parameters as Ellipsim. The principal difference is in connectivity. The DFN method creates the fracture network connectivity via an iterative process in which not connected clusters of fractures are removed. The permeability fields were generated with FCM using three different methods and the same fracture data set loosely based on the data from an existing site in granite rocks. A few examples of transport simulations are provided to demonstrate the major findings of the comparison.
Modeling of heat extraction in Enhanced Geothermal Systems is presented. The study builds on recent studies on the use of directional wells to improve heat transfer between doublet injection and production wells. The current study focuses on the influence of fracture orientation on production temperature in deep low permeability geothermal systems, and the effects of directional drilling and separation distance between boreholes on heat extraction. The modeling results indicate that fracture orientation with respect to the well-pair plane has significant influence on reservoir thermal drawdown. The vertical well doublet is impacted significantly more than the horizontal well doublet.
A set of 16 geologic disposal concepts is described in sufficient detail for rough-order-of-magnitude repository cost estimates, for disposal of spent nuclear reactor fuel in generic crystalline, argillaceous, and salt host geologic media. The description includes total length, diameter, and volume for all underground shafts, ramps, drifts and large-diameter borings. Basic types of ground support are specified. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel, but concepts are described in terms of modular panels each containing 10,000 MT. Waste packaging is described, and the materials and outer dimensions for disposal overpacks are given. The manner of emplacement is specified, with any additional fixturing, lining, buffer materials, and backfill needed. Thermal limits for waste package emplacement or repository closure are given, as appropriate for enclosed and open emplacement modes, respectively. The 16 disposal concepts are based on disposal concept studies performed for the Used Fuel Disposition R&D program between 2011 and 2014, and the accompanying engineering analyses. Revision 1 results from review and checking of repository spacings, drift length estimates, repository plan area estimates, closure power limits, and estimates of fuel age at closure.