Analysis on Thermal-hydraulic Characteristics of Lead-bismuth Reactor System LESMOR under Vehicle-mounted Motion

被引：0
|
作者：
Lu D. [1 ]
Wei S. [2 ]
Wang C. [1 ]
Tian W. [1 ]
Su G. [1 ]
Qiu S. [1 ]
机构：
[1] School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an
[2] China Nuclear Power Technology Research Institute Co., Ltd., Shenzhen
来源：
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | 2023年 / 57卷 / 01期
关键词：
lead-bismuth fast reactor; thermal-hydraulic and safety analysis; vehicle-mounted condition;
D O I：
10.7538/yzk.2022.youxian.0035
中图分类号：
学科分类号：
摘要：
The construction of small modular reactor design is mostly for commercial power generation, and there are certain limitations in maneuverability and flexibility. In order to meet the demand of energy supply for special tasks, the vehicle-mounted very small modular reactor with more maneuverability arises at the right moment. In practical operation, vehicle-mounted reactor system will have multi-degree-of-freedom space movement, and the influence of vehicle movement on the reactor system is not clear at present. In this paper, the complex motion forms of the vehicle-mounted nuclear energy system running on the road were simplified into three types of operation forms. Based on the fluid force analysis in the non-inertial system, the additional force of vehicle-mounted motion conditions was derived from the momentum conservation equation, and a mathematical and physical model of vehicle-mounted motion conditions was established. The influence of additional force under motion condition on liquid lead-bismuth eutectic was analyzed reasonably, and corresponding additional force model was added into LETHAC program of transient thermal-hydraulic analysis program of lead-bismuth fast reactor to analyze thermal-hydraulic safety characteristics of LESMOR system under motion condition. The results show that the natural circulation flow of LESMOR system decreases significantly and the core cladding temperature and fuel temperature increase significantly under ramp driving, vertical vibration, pitching vibration and vertical coupled pitching vibration. Slope has little influence on the natural circulation of the system, and the change amplitude of system flow rate and temperature is small. The vertical vibration and pitching vibration will decrease the time domain average flow rate of the system and increase the time domain average temperature of the core outlet coolant. Due to the arrangement of the system, the pitching vibration has more influence on the heat exchanger than on the core. When the pitching vibration and vertical vibration arc coupled, the average flow rate of the system in time domain decreases, and the vertical vibration plays a dominant role in the core flow rate. Under the limit condition, the transient drop of natural circulation flow of the system is not more than 30%, and the transient temperature of cladding and fuel is up to 411. 3 °C and 442. 4 °C, and the rise is not more than 4 °C, which is far below the safety limit of transient operation. The LESMOR system maintains stable and safe operation in both normal and possible driving limits. This paper provides a method to analyze thermal safety characteristics of reactor system under vehicle-mounted motion, which has certain reference significance. © 2023 Atomic Energy Press. All rights reserved.
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页码：111 / 123
页数：12
相关论文
共 14 条
  • [1] STERBENTZ J W., Special purpose nuclear reactor (5 MW) for reliable power at remote sites assessment report, (2017)
  • [2] STAUFF N., Neutronic design and analysis of the holos-quad concept, (2019)
  • [3] YU H, HARTANTO D, MOON J, Et al., A conceptual study of a supercritical CO2-cooled micro modular reactor, Energies, 8, 12, pp. 13938-13952, (2015)
  • [4] GAO Puzhen, PANG Fengge, Mathematical model of primary coolant in nuclear power plant affected by ocean conditions, Journal of Harbin Engineering University, 18, 1, pp. 26-29, (1997)
  • [5] CHANG W, WANG S, HAO W, Et al., Investigation of flow pulsation characteristic in single-phase forced circulation under rolling motion, Annals of Nuclear Energy, 64, pp. 50-56, (2014)
  • [6] WEI S, MA W, WANG C, Et al., Development and validation of transient thermal-hydraulic evaluation code for a lead-based fast reactor, International Journal of Energy Research, (2021)
  • [7] WANG C, WEI S, TIAN W, Et al., Sub-channel analysis for Pb-Bi-cooled direct contact boiling water fast reactor, International Journal of Energy Research, 42, 8, pp. 2643-2654, (2018)
  • [8] WEI S, WANG C, TIAN W, Et al., Transient thermal-hydraulic evaluation of lead-bismuth fast reactor by coupling sub-channel and system analysis codes, Nuclear Engineering and Design, 337, pp. 228-235, (2018)
  • [9] LIANG Y, ZHANG D, CHEN Y, Et al., An experiment study of pressure drop and flow distribution in subchannels of a 37-pin wire-wrapped rod bundle, Applied Thermal Engineering, 174, (2020)
  • [10] WU YW, LI X, YU X, Et al., Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor, Progress in Nuclear Energy, 68, pp. 65-78, (2013)
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