The traditional solution of the coupled neutronics/ thermal-hydraulics problems has typically been performed by solving the individual field separately and then transferring information between each other. In this paper, full implicit integrate solution to the coupled neutronics/ thermal-hydraulic problem is investigated. There are two advantages compared with the traditional method, which are high temporal accuracy and stability. The five equations of single-phase flow, the solid heat conduction and the neutronics are employed as a simplified model of a nuclear reactor core. All these field equations are solved together in a tightly coupled, nonlinear fashion. Firstly, Newton-based method is employed to solve nonlinear systems due to its local second-order convergence rate. And then the Krylov iterative method is used to solve the linear systems which are from the Newton linearization. The two procedures above are the so-called Newton-Krylov method. Furthermore, in order to improve the performance of the Krylov method, physics-based preconditioner is employed, which is constructed by the physical insight. Finally, several Newton-Krylov solution approaches are carried out to compare the performance of the coupled neutronics / thermal-hydraulic equations.
Skip Nav Destination
2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5580-5
PROCEEDINGS PAPER
Full Implicit Integrate Solution to the Coupled Neutronics/Thermal-Hydraulics Problems
Han Zhang
Tsinghua University, Beijing, China
Fu Li
Tsinghua University, Beijing, China
Paper No:
ICONE21-16159, V003T10A036; 5 pages
Published Online:
February 7, 2014
Citation
Zhang, H, & Li, F. "Full Implicit Integrate Solution to the Coupled Neutronics/Thermal-Hydraulics Problems." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 3: Nuclear Safety and Security; Codes, Standards, Licensing and Regulatory Issues; Computational Fluid Dynamics and Coupled Codes. Chengdu, China. July 29–August 2, 2013. V003T10A036. ASME. https://doi.org/10.1115/ICONE21-16159
Download citation file:
26
Views
Related Proceedings Papers
Related Articles
Thermal-Hydraulics Program in Support of Canadian SCWR Concept Development
ASME J of Nuclear Rad Sci (January,2018)
Special Section: Pressurized Heavy Water Reactors Safety
ASME J of Nuclear Rad Sci (April,2017)
Methodology for Calculating Minor Radioactive Releases From VVER 1000 Using TRACE Code
ASME J of Nuclear Rad Sci (April,2021)
Related Chapters
Steady State Reactor Thermal Hydraulics
Nuclear Reactor Thermal-Hydraulics: Past, Present and Future
A General Thermal Hydraulics Uncertainty Analysis Methodology (PSAM-0422)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Front Matter
Nuclear Reactor Thermal-Hydraulics: Past, Present and Future