Session: 02-13: Structural Evaluation, Performance Assessment, Multiphysics Coupling - III

Paper Number: 134828

134828 - Neutronic and Thermal-Hydraulic Analysis for a Small Liquid-Solid Dual-Fuel Reactor With Different Fuel Types 

Abstract: 

Small molten salt reactor (SMSR) is an advanced nuclear energy system that uses molten salt as fuel and coolant. Compared with conventional nuclear reactors, SMSRs have the advantages of high safety, high thermal and electrical efficiency, low research and development costs, low power generation prices and long core life. In this study, a small molten salt reactor using cross-spiral fuel was innovatively proposed. The small molten salt reactor adopts UC with U235 enricment 45.89% as fuel, and the molten salt coolant is ⁷LiF-KF-UF₄ (49.55-49.55-0.9 mol%) fluoride salt with U235 enricment 30%. The reactor core contains 61 fuel rods, and its radius and height are respectively 25 cm and 43.2 cm. BeO served as the reflector, and the spectral shift absorber material is Gd₂O₃, and a control drum containing B₄C was employed to control reactivity. Furthermore, the neutronic and thermal-hydrarulic characteristic were analyzed and compared with a conventional cylindrical rod fuel using a Monte Carlo code OpenMC deveopled by MIT and a CFD software, respectively. In neutronics, the cross-spiral fuel with different spiral angles was not considered since the OpenMC can’t model it. The neutronic results show that the two fuel types have the similar fast neutron spectrum, and the k_eff of the cross fuel is slightly smaller than that of the cylindrical rod fuel. In steady state, the k_eff is 1.042 for the rod fuel and 1.038 for the cross fuel, and about 100 pcm difference can be found for those two fuel types. The reactivity coefficients of the two fuel configurations are both negative, and the fuel temperature reactivity coefficient is -1.132 pcm/K for the rod fuel and -1.156 pcm/K for the cross-spiral fuel. The criticality analysis of reactor under the water flooding and sand burial accidents were performed, and the rector with both two fuel types would not return to the criticality. Moreover, in thermal-hydraulic analysis, the inlet temperature and inlet flow rate of the molten salt adopted 873 K and 0.1 m/s for all coolant channel ignoring the radial flow rate distribution. The themal-hydraulic analysis shows that the maximum temperature of the reactor core with the rod fuel is 1435.4 K, and the maximum temperature of the reactor core with the spiral cross fuel is 1231.21 K. The maximum temperature of the core is 14.2% lower than that of the rod fuel. This is because the coolant flow in the core with rod fuel rods is approximately parallel to the rod bundles, while the spiral-cross fuel resulting stronger turbulence. Thus, the spiral cross-shaped fuel rods can well reduce the maximum temperature during core operation and have a better effect on the temperature levelling of the reactor.

Presenting Author: Kun Zhuang Nanjing University of Aeronautics and Astronautics

Presenting Author Biography: Zhuang Kun, associate professor, Nanjing University of Aeronautics and Astronautics. Visit scholar of University of Michigan at 2015.

Authors:

Zhichao Qiu Nanjing University of Aeronautics and Astronautics
Kun Zhuang Nanjing University of Aeronautics and Astronautics
Yongzhan Wang Nanjing University of Aeronautics and Astronautics
Xiaoyu Wang Nanjing University of Aeronautics and Astronautics
Lina Deng Nanjing University of Aeronautics and Astronautics
Yingzheng Wang Nanjing University of Aeronautics and Astronautics
Sipeng Wang Nanjing University of Aeronautics and Astronautics