Abstract
Application of the Virtual Environment for Reactor Applications (VERA) to BWR analysis is assessed in this paper by comparing results to those calculated using other widely-used modeling tools, namely the U.S. Nuclear Regulatory Commission's PARCS/PATHS and the Serpent Monte Carlo particle transport code. Additionally, VERA is used to calculate 3-D temperature and fast neutron flux distributions in silicon carbide (SiC) fiber-reinforced, SiC matrix composite (SiC/SiC) BWR channel boxes, which are being studied as an Accident Tolerant Fuel core structural material concept. The code-to-code comparisons were favorable, and the SiC/SiC channel box evaluation demonstrates the many advanced modeling features VERA provides while also highlighting the non-uniformity in fast neutron flux distributions that can play a role in potential SiC/SiC channel box deformation. Traditional BWR analysis tools do not have the calculation fidelity necessary for coupled assessment of flux and temperature gradients in a SiC/SiC channel box. VERA is a state-of-the-art modeling environment that was developed to increase the safety and economic competitiveness of nuclear power through improved modeling accuracy. While VERA has already been deployed in the nuclear industry for PWR applications, the current study is a vital initial step in the extensive development, validation, and verification that VERA must go through to be useful for BWR applications.
| Original language | English |
|---|---|
| Article number | 110732 |
| Journal | Nuclear Engineering and Design |
| Volume | 365 |
| DOIs | |
| State | Published - Aug 15 2020 |
Bibliographical note
Publisher Copyright:© 2020 Elsevier B.V.
Funding
The authors would like to thank the Department of Energy Office of Nuclear Energy Advanced Fuels Campaign and the Consortium for Advanced Simulation of Light Water Reactors for their support of this work. The authors would also like to thank Dr. Andrew Ward from the University of Michigan for providing a pre-released version of GenPMAXS. While not used in this analysis, the willingness to collaborate is greatly appreciated. All computational simulations were performed at the Oak Ridge National Laboratory. This manuscript has been co-authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan). The authors would like to thank the Department of Energy Office of Nuclear Energy Advanced Fuels Campaign and the Consortium for Advanced Simulation of Light Water Reactors for their support of this work. The authors would also like to thank Dr. Andrew Ward from the University of Michigan for providing a pre-released version of GenPMAXS. While not used in this analysis, the willingness to collaborate is greatly appreciated. All computational simulations were performed at the Oak Ridge National Laboratory.
| Funders | Funder number |
|---|---|
| Department of Energy Office of Nuclear Energy Advanced Fuels Campaign | |
| U.S. Department of Energy | |
| University of Michigan |
Keywords
- Accident Tolerant Fuel
- Boiling water reactor
- Code-to-code comparison
- High-fidelity
- Multiphysics