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Tutorial: Automated Assembly-Specific/As-loaded and Design-basis Spent Nuclear Fuel and Related Systems Characterizations Using UNF-ST&DARDS
Thursday 1:30-5:00 PM
The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) is being developed at Oak Ridge National Laboratory for integrating spent nuclear fuel (SNF) management through its final disposition. UNF-ST&DARDS provides a database for storing and preserving SNF data and streamlines various analyses using the data from the database for time-dependent characterization of SNF and related systems (e.g., dry storage system). When basic information about the SNF and the cask system is provided, the data relationships defined in UNF-ST&DARDS allow inputs to the respective codes (e.g., SCALE) to be built autonomously. This tutorial will include (1) discussions on various data need for realistic SNF characterizations and demonstration of the data import processes in UNF-ST&DARDS, (2) assembly-specific, time-dependent depletion and decay analyses, (3) as-loaded (using actual cask loading maps) criticality and shielding analyses of currently loaded casks for storage, transportation, and disposal (over disposal time periods), (4) discussion of as-loaded analyses to support licensing/certification of dry SNF systems, (5) misload analysis methodology to support as-loaded criticality analysis, and (6) design-basis criticality and shielding analyses using user-defined loading patterns and fuel assembly types.
The tutorial/demonstration is open to all registered meeting attendees who signed up at the time of registration. Laptops are not required, but participants with valid licenses for UNF-ST&DARDS and SCALE 6.2.2/SCALE 6.2.3 installed on their computer (windows machine) can follow the demonstrations hands-on.
PFLOTRAN IHLRWM Short Course
Thursday, April 18 Afternoon
PFLOTRAN is an open source, massively-parallel reactive multiphase flow and transport code being developed by researchers from around the world to simulate problems of varying complexity, from simple 1D transport to large 3D multiphase flow and biogeochemical reaction in heterogeneous porous media. The code is founded upon the parallel PETSc framework and designed to run on computers ranging from laptops to supercomputers. PFLOTRAN is being employed by the US Department of Energy’s (DOE) Spent Fuel and Waste Sciences and Technologies (SFWST) Campaign to simulate thermal, hydrologic, and chemical processes associated with deep geologic disposal of radioactive waste. It is an integral component of the DOE’s geologic disposal safety assessment (GDSA) framework (pa.sandia.gov). This half-day short course will demonstrate PFLOTRAN capability within the context of radioactive waste management and include presentation of underlying theory, demonstration of simulation execution, and visualization of results. Participants who would like to install a virtual machine for running PFLOTRAN demonstration problems are encouraged to bring a laptop.
MAVRIC shipping cask analysis using ORIGEN source terms (4 hrs)
Sunday, April 14
One of the unique features of the SCALE code system is the flexibility of assembling different SCALE codes or sequences to solve complex problems. Transportation and storage of spent fuel require a computational tool set to characterize both the spent fuel source terms and the doses for containers used to transport or store the fuel. Spent fuel is a complex neutron and photon source that can be well characterized using the ORIGEN code in SCALE. Additionally, ORIGEN can be used to characterize the radioactive sources resulting from activation of non-fissile materials and components in a nuclear reactor, such as the pressure vessel. The variety of source terms generated with ORIGEN can be used for shielding analyses with the MAVRIC sequence. MAVRIC can estimate particle fluxes and dose rates outside of containers, to ensure that the safety requirements for transportation, storage and ultimate disposal of spent fuel or activated materials are met.
This workshop will provide an introduction to ORIGEN and will demonstrate how to use the ORIGAMI tool for quick calculations of spent fuel sources. This will be followed by an introduction to MAVRIC/Monaco and the automated variance reduction techniques CADIS and FW-CADIS. Finally, the sources generated by ORIGAMI will be used as a source in a MAVRIC/Monaco dose analysis outside of a shipping cask. Registered SCALE users are welcome to bring their laptop and follow along.
Cyclus fuel cycle simulator
Sunday, April 14
This tutorial will demonstrate how to use Cyclus to simulate both simple and complex nuclear fuel cycle scenarios. It is appropriate for all levels of expertise, and attendees should bring a laptop if they intend to follow along with the exercises. Cyclus (fuelcycle.org) is the next-generation agent-based nuclear fuel cycle simulator, providing flexibility to users and developers through a dynamic resource exchange solver and plug-in, user-developed agent framework. The goal of Cyclus is to enable a broad spectrum of fuel cycle simulation while providing a low barrier to entry for new users and agent developers. Cyclus engages with potential module developers and encourages them to join a vibrant community in an expanding ecosystem. Users and developers are always welcome and encouraged to use or contribute to the Cyclus project.