The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS
Abstract
The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, the capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.
- Authors:
-
- Univ of Michigan, Ann Arbor, MI (United States)
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- U.S. Nuclear Regulatory Commission (NRC-RES), Washington, DC (United States)
- Publication Date:
- Research Org.:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Org.:
- USDOE: USNRC
- OSTI Identifier:
- 1261441
- Grant/Contract Number:
- AC05-00OR22725; NRC-04-10-149; NRC-HQ-60-13-D-0018
- Resource Type:
- Accepted Manuscript
- Journal Name:
- Nuclear Technology
- Additional Journal Information:
- Journal Volume: 190; Journal Issue: 3; Journal ID: ISSN 0029-5450
- Publisher:
- Taylor & Francis - formerly American Nuclear Society (ANS)
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
Citation Formats
Ward, Andrew, Downar, Thomas J., Xu, Y., March-Leuba, Jose A, Thurston, Carl, Hudson, Nathanael H., Ireland, A., and Wysocki, A. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS. United States: N. p., 2015.
Web. doi:10.13182/NT14-79.
Ward, Andrew, Downar, Thomas J., Xu, Y., March-Leuba, Jose A, Thurston, Carl, Hudson, Nathanael H., Ireland, A., & Wysocki, A. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS. United States. https://doi.org/10.13182/NT14-79
Ward, Andrew, Downar, Thomas J., Xu, Y., March-Leuba, Jose A, Thurston, Carl, Hudson, Nathanael H., Ireland, A., and Wysocki, A. Wed .
"The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS". United States. https://doi.org/10.13182/NT14-79. https://www.osti.gov/servlets/purl/1261441.
@article{osti_1261441,
title = {The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS},
author = {Ward, Andrew and Downar, Thomas J. and Xu, Y. and March-Leuba, Jose A and Thurston, Carl and Hudson, Nathanael H. and Ireland, A. and Wysocki, A.},
abstractNote = {The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, the capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.},
doi = {10.13182/NT14-79},
journal = {Nuclear Technology},
number = 3,
volume = 190,
place = {United States},
year = {Wed Apr 22 00:00:00 EDT 2015},
month = {Wed Apr 22 00:00:00 EDT 2015}
}
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