Modeling and simulation has a long history with researchers and scientists exploring nuclear energy technologies. In fact, the existing fleet of currently operating reactors was licensed with computational tools that were produced or initiated in the 1970s. Researchers and scientists in the Department of Energy are developing new tools to predict the performance, reliability and economics of advanced nuclear power plants. The new computational tools will allow researchers to explore in ways never before practical, at the level of detail dictated by the governing phenomena, all the way from important changes in the materials of a nuclear fuel pellet to the full-scale operation of a complete nuclear power plant.
Advanced Reactor Technology Simulation Modelling and Framework
Nuclear Systems Analysis Code
Neutronics and Shielding Analysis Enabling Nuclear Technology Advancements
The Virtual Environment for Reactor Applications (VERA) toolkit being developed by the Consortium for Advanced Simulation of Light water reactors (CASL) that provides an environment for simulating light water reactors (LWRs) to address industry-driven challenges. The tools in VERA focus on steady-state and transient analysis of nuclear fuel, reactor core, and vessel with integrated neutronics, thermal-hydraulics, chemistry and fuel performance. The physics tools in VERA also provide capabilities that are relevant to advanced reactor concepts.
SHARP is a suite of high-fidelity reactor simulation tools comprised of thermal hydraulic, neutronics, and structural mechanics modules as well as various supporting tools that collectively form the Reactors Product Line of the NEAMS ToolKit. Each module can be utilized as a standalone code component or as part of an integrated analysis. SHARP was developed with technology neutral simulation capabilities in mind so that it could extend to various types of advanced reactors. Initial efforts focus on development of tools for analysis of performance and safety of sodium-cooled fast reactors (SFR).
The SAS4A/SASSYS-1 code system is designed to perform deterministic analysis of design basis and beyond-design basis accidents in fast reactor plants. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, neutronic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by loss of coolant flow, loss of heat rejection, or reactivity insertion.
The Argonne Reactor Computation (ARC) code system is used for reactor core design and fuel cycle analysis. It is also used to generate neutronics and thermal-hydraulic information for plant safety analysis. While designed primarily for fast reactors, it is applicable more broadly to a wide range of reactor technologies. The ARC system provides capabilities for multi-group cross-section generation; neutron diffusion and transport calculation; fuel depletion modeling; in-core and ex-core fuel cycle analysis; and calculation of reactivity effects of perturbations in reactor conditions. It is integrated with steady-state sub-channel thermal hydraulics analysis, and provides input for plant transient and safety analyses performed using the the SAS4A/SASSYS-1 code.