摘要
Abstract
The lead-cooled fast reactor(LFR),as one of the generation Ⅳ nuclear systems,attracts considerable attention due to its excellent thermophysical and neutronic characteristics.With increasing demands for core safety and economic performance,accurate neutronics/thermal-hydraulics coupling analysis becomes increasingly important for evaluating reactor behavior.The Monte Carlo method,characterized by its use of continuous-energy cross sections and flexible geometric modeling,enables detailed and reliable neutron transport simulations.Meanwhile,in the field of thermal-hydraulic analysis,subchannel codes achieve an optimal balance between computational precision and efficiency,making them suitable for pin-level core simulations.This study aims to develop a high-resolution internal coupling framework through API integration and in-memory data exchange techniques,which enables the deep exploration of complex multi-physics feedback mechanisms in LFRs under both steady-state and transient conditions.The primary objective is to satisfy the requirements for detailed neutronics/thermal-hydraulics coupling calculations in LFRs,thereby providing a reliable analysis tool for core design and safety assessment.The OpenMC/KMC-FBc/PK internal coupling system was established in this study by integrating the Monte Carlo neutron transport code OpenMC,the pin-level full-core subchannel program KMC-FBc,and a point kinetics(PK)model.To overcome the efficiency bottlenecks of cross-language communication,the Pybind11 library was employed to encapsulate the Monte Carlo module,enabling the C++main control program to invoke application programming interfaces and perform direct in-memory data exchange.In this framework,the PK model was implemented to solve for transient power evolution,while the Monte Carlo module provided the detailed three-dimensional spatial power distribution,and the subchannel program was utilized to execute the thermal-hydraulic response during transients.Consequently,this approach maintained the high spatial resolution of both the Monte Carlo and subchannel codes while ensuring computational efficiency for transient solutions.The system was verified through code-to-code comparisons using a small-scale LFR model designed by Shanghai Jiao Tong University and the SNCLFR-100 reactor model developed by the University of Science and Technology of China.Furthermore,the system was applied to the NCLFR-Oil reactor for comprehensive steady-state and unprotected loss of heat sink(ULOHS)analysis.The results indicate that the spatial power distributions and key thermal-hydraulic parameters calculated by OpenMC/KMC-FBc/PK show excellent agreement with benchmark codes,verifying its accuracy and reliability in fine-grained coupled analysis for LFRs.Steady-state and transient analyses conducted on the NCLFR-Oil core demonstrate that while the coupling effects in LFRs have a less significant impact on power distribution compared to pressurized water reactors,they markedly influence key thermal-hydraulic parameters.Under ULOHS conditions,the peak temperatures calculated by OpenMC/KMC-FBc/PK are higher than those from the system-level code ATHLET,providing a more accurate reflection of local hot-spot characteristics.Overall,the developed neutronics/thermal-hydraulics coupling program accurately describes the interactions between neutron transport and thermal-hydraulics.It proves suitable for both steady-state and transient LFR analysis,serving as a reliable tool for the performance evaluation and safety assessment of LFRs.关键词
铅冷快堆/核热耦合/OpenMC/pin-by-pinKey words
lead-cooled fast reactor/neutronics/thermal-hydraulics coupling/OpenMC/pin-by-pin分类
能源科技
