The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS
- 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)
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.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE: USNRC
- Grant/Contract Number:
- AC05-00OR22725; NRC-04-10-149; NRC-HQ-60-13-D-0018
- OSTI ID:
- 1261441
- Journal Information:
- Nuclear Technology, Vol. 190, Issue 3; ISSN 0029-5450
- Publisher:
- Taylor & Francis - formerly American Nuclear Society (ANS)Copyright Statement
- Country of Publication:
- United States
- Language:
- English
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