Session: 04-07: SMRs, Advanced Reactors and Fusion

Paper Number: 135246

135246 - Thermo-Neutronics Coupled Simulation of a Heat Pipe Reactor Based on Comsol 

Abstract: 

As an advanced small nuclear reactor, the heat pipe reactor possesses several advantages, including high energy density, long operational lifetime, compact size, and strong adaptability to various environments, making it an optimal choice for specialized energy needs in future applications, such as deep-sea and deep-space domains. In this study, we developed a code system using OpenMC/COMSOL for neutron and thermodynamic simulations. The continuous-energy Monte Carlo code, OpenMC, was employed to generate homogenized cross-section databases, offering significant modeling flexibility compared to traditional deterministic lattice transport codes. The generated multi-group cross-sections from OpenMC were utilized in COMSOL for the coupled neutron and thermodynamic simulations of the entire core. To validate the OpenMC/COMSOL code system, benchmark problems for pressurized water reactors were computed, and the results of the "two-step" scheme for neutron physics were compared with full-core Monte Carlo neutron results. In addition, in order to study the applicability of hydrothermal coupling in the neutron physics model of heat pipe reactors, typical fine models of heat pipe reactor components and neutron physics calculation models were established. The neutron physics calculation models were verified under different energy group numbers and uniform regions. The results demonstrated good agreement for multiplication factors and power distributions. To demonstrate the capability of the OpenMC/COMSOL code system in predicting transient characteristics of the heat pipe reactor core, a comprehensive full-core coupled neutron and thermodynamic simulation was performed on a representative simplified heat pipe reactor core, investigating the transient phenomena caused by heat pipe failure scenarios. The research results indicate that the utilization of the cross-section library generated by OpenMC enables the capability for steady-state analysis and core design. The results obtained from COMSOL exhibit good overall agreement with respect to multiplication factors and power distribution. Regarding the heat pipe reactor, both energy group partitioning and region division significantly influence the calculation accuracy. Considering the trade-off between computation accuracy and calculation resource consumption, it is recommended to adopt the two-group cross-section with SPH (Simplified P1) correction for diffusion calculations and use a fine model to determine the internal power distribution. The analysis of typical transient characteristics of heat pipe reactors shows that the OpenMC/COMSOL code significantly improves computational efficiency, and the transient calculation reflects the self stability characteristics of the core, which proves that the code effectively captures the local effects of transient conditions on the heat transfer performance and structural mechanics inside the heat pipe reactor

Presenting Author: Jingyu Nie Xi'an Jiaotong University

Presenting Author Biography: Main research direction in thermal hydraulic engineering and safety analysis of micro reactors

Authors:

Jingyu Nie Xi'an Jiaotong University
Binqian Li Xi'an Jiaotong University
Yingwei Wu Xi'an Jiaotong University
Jing Zhang Xi'an Jiaotong University
Guoliang Zhang China Nuclear Power Technology Research Institute
Qisen Ren China Nuclear Power Technology Research Institute
Yanan He Xi'an Jiaotong University
Guanghui Su Xi'an Jiaotong University