Abstract
The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics (T/H), and fuel temperature components with an isotopic depletion capability. The neutronics capability is based on the Michigan Parallel Characteristics Transport Code (MPACT), a three-dimensional whole-core transport code. The T/H and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to the departure from nucleate boiling (DNB) ratio at the most limiting point of a postulated pressurized water reactor main steam line break event initiated at the hot zero power, either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power, where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady-state reactor core response under the main steam line break accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.
| Original language | English |
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| Title of host publication | Computational Fluid Dynamics (CFD) and Coupled Codes; Decontamination and Decommissioning, Radiation Protection, Shielding, and Waste Management; Workforce Development, Nuclear Education and Public Acceptance; Mitigation Strategies for Beyond Design Basis Events; Risk Management |
| Publisher | American Society of Mechanical Engineers (ASME) |
| ISBN (Electronic) | 9780791850046 |
| DOIs | |
| State | Published - 2016 |
| Event | 2016 24th International Conference on Nuclear Engineering, ICONE 2016 - Charlotte, United States Duration: Jun 26 2016 → Jun 30 2016 |
Publication series
| Name | International Conference on Nuclear Engineering, Proceedings, ICONE |
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| Volume | 4 |
Conference
| Conference | 2016 24th International Conference on Nuclear Engineering, ICONE 2016 |
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| Country/Territory | United States |
| City | Charlotte |
| Period | 06/26/16 → 06/30/16 |
Funding
This paper contains results of research supported by the Consortium for Advanced Simulation of Light Water Reactors (www.casl.gov), an Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors (http://www.energy.gov/hubs) under U.S. Department of Energy (DOE) Contract No. DE-AC05-00OR22725. This research also used resources of the Oak Ridge Leadership Computing Facility at the Oak Ridge National Laboratory, which is supported by the DOE Office of Science under Contract No. DE-AC05-00OR22725.