热管冷却反应堆固态堆芯热力耦合性能分析及结构优化

岳明楷; 金浩; 刘桐蔚; 李文杰; 黄永忠; 恽迪; 周进雄

doi:10.7538/yzk.2022.youxian.0630

热管冷却反应堆固态堆芯热力耦合性能分析及结构优化

西安交通大学航天航空学院机械振动与强度国家重点实验室,陕西西安710049
中国核动力研究设计院,四川成都610065
西安交通大学能源与动力学院,陕西西安710049

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- 刊出日期: 2023-02-19

Thermal-mechanical Coupling Analysis and Structure Optimization of Solid Core of Heat-pipe-cooled Reactor

State Key Laboratory for Strength and Vibration of Mechanical Structures, School of Aerospace, Xi’an Jiaotong University, Xi’an 710049, China
Nuclear Power Institute of China, Chengdu 610065, China
School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an 710049, China

摘要

摘要: 热管冷却反应堆的固态堆芯具有模块化、无流动回路等诸多优势，受到诸多学者关注。固态堆芯在工作寿期内涉及复杂的热力耦合行为。本文基于ABAQUS对固态堆芯工作性能进行了热力耦合分析。结合子程序的二次开发，综合考虑了间隙传热模型及材料在极端工况下的蠕变、肿胀等行为。探究了各部件之间的相互作用关系。基于有限元仿真对固态堆芯尺寸进行参数优化，使用深度神经网络建立代理模型，使用NSGA-Ⅱ算法获得了pareto前沿解集。相较于初始设计参数，优化后的堆芯最高温度下降8.44%，最大应力下降34.43%，改善了固态堆芯的性能。
- 热管冷却反应堆固态堆芯 ,
- 间隙热传导 ,
- 热力耦合 ,
- 深度神经网络 ,
- 尺寸优化
Abstract: Heat-pipe-cooled reactors offer many advantages, such as modularity and no flow loops. In recent years, it has attracted the attention of many scholars. Solid cores involve complex thermal-mechanical coupling behavior over their operating lifetime. The deformation of solid core structure has significant influence on its heat transfer performance. This paper introduced the gap heat transfer model by Ross and Stoute, which is an important part of the heat transfer process in the core. The creep and swelling behavior of materials used in solid core under high temperature and irradiation environment was briefly described. In this paper, a simplified finite element model of solid core was developed based on the secondary development of ABAQUS by programming subroutines, which considered the effect of gap heat transfer, creep and swelling behavior in the core. Based on the Ross-Stoute gap heat transfer model, the subroutine GAPCON was developed to simulate the gap heat transfer behavior in the core. The subroutine CREEP was developed to simulate the complex mechanical behavior of core components under different temperatures and radiation environments. Typical characteristics of temperature and stress distribution in each component during the normal operation period of 5 years were obtained. The thermal-mechanic coupling behaviors of solid core and the interaction between components were analyzed. The high temperature area appears in the center of the fuel pellets, and the temperature drops along the radial direction of the fuel pellets. This distribution of temperature field results in the inner extrusion and the outer tension of fuel pellets. The annular tensile stress region appears obviously outside the fuel pellets. These temperature and stress distribution characteristics can partly explain the radial cracking behavior of cylindrical fuel pellets. Furthermore, the designed size of the solid core was optimized based on finite element and machine learning to obtain lower temperature and stress peaks. The parametric modeling technique of ABAQUS was used to generate 5 800 sets of data, and the sample database was established for the training and testing of deep neural network machine learning. A surrogate model of core design parameters, maximum temperature and Mises stress in the reactor was established. The surrogate model had good predictive accuracy. Based on the surrogate model, pareto front was obtained by NSGA-Ⅱ algorithm and the optimized design parameters were obtained. The results of core temperature and stress optimized by surrogate model were verified by finite element method. Compared with the initial design, the maximum temperature decreases by 8.44% and the maximum stress decreases by 34.43%. The performance of the core is improved.
- solid core of heat-pipe-cooled reactor ,
- gap conductance ,
- thermal-mechanical coupling ,
- deep neural network ,
- size optimization

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[1]	余红星,马誉高,张卓华,等. 热管冷却反应堆的兴起和发展[J]. 核动力工程,2019,40(4):1-8. YU Hongxing, MA Yugao, ZHANG Zhuohua, et al. Initiation and development of heat pipe cooled reactor[J]. Nuclear Power Engineering, 2019, 40(4): 1-8(in Chinese).
[2]	GROVER G M, COTTER T P, ERICKSON G F. Structures of very high thermal conductance[J]. Journal of Applied Physics, 1964, 35(6): 1990-1991.
[3]	RANKEN W A, HOUTS M G. Heat pipe cooled reactors for multi-kilowatt space power supplies[R]. US: LANL, 1995.
[4]	HOUTS M G, POSTON D I, RANKEN W A. Heatpipe space power and propulsion systems[C]∥AIP Conference Proceedings. US: American Institute of Physics, 1996: 1155-1162.
[5]	MCCLURE P R, POSTON D I, DIXON D D. Final results of demonstration using flattop fissions (DUFF) experiment[R]. US: LANL, 2012.
[6]	VANDYKE M, MARTIN J. Non-nuclear testing of reactor systems in the early flight fission test facilities (EFF-TF)[C]∥2004 International Congress on Advances in Nuclear Power Plants (ICAPP 2004). [S. l.]: [s. n.], 2004.
[7]	MCCLURE P R, POSTON D I, GIBSON M A, et al. Kilopower project: The KRUSTY fission power experiment and potential missions[J]. Nuclear Technology, 2020, 206: 1-12.
[8]	王傲,申凤阳,胡古,等. 热管空间核反应堆电源的研究进展[J]. 核技术,2020,43(6):9-15. WANG Ao, SHEN Fengyang, HU Gu, et al. A survey of heatpipe space nuclear reactor power supply[J]. Nuclear Techniques, 2020, 43(6): 9-15(in Chinese).
[9]	孙浩,王成龙,刘逍,等. 水下航行器微型核电源堆芯设计[J]. 原子能科学技术,2018,52(4):646-651. SUN Hao, WANG Chenglong, LIU Xiao, et al. Reactor core design of micro nuclear power source applied for underwater vehicle[J]. Atomic Energy Science and Technology, 2018, 52(4): 646-651(in Chinese).
[10]	袁乃驹. 核工程中采用热管的探讨[J]. 核动力工程, 1980,1(3):52-54,44. YUAN Naiju. Discussion on the application of heat pipes in nuclear engineering[J]. Nuclear Power Engineering, 1980, 1(3): 52-54, 44(in Chinese).
[11]	田智星,刘逍,王成龙,等. 高温钾热管稳态运行传热特性研究[J]. 原子能科学技术,2020,54(10):1771-1778. TIAN Zhixing, LIU Xiao, WANG Chenglong, et al. Study on heat transfer performance of high temperature potassium heat pipe at steady state[J]. Atomic Energy Science and Technology, 2020, 54(10): 1771-1778(in Chinese).
[12]	ROSS A M, STOUTE R L. Heat transfer coefficient between UO₂ and zircaloy-2[R]. Ontario: Atomic Energy of Canada Limited, 1962.
[13]	KENNARD E H. Kinetic theory of gases[M]. New York: McGraw-hill, 1938.
[14]	HAGRMAN D T, ALLISON C M, BERNA G A. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4[R]. US: NRC, 1995.
[15]	GILBERT E R, BLACKBURN L D. Creep deformation of 20 percent cold worked type 316 stainless steel[J]. J Eng Mater Technol, 1977, 99(2): 168-180.
[16]	LI R, ZHOU Y. High temperature creep properties of UO₂ fuel pellets manufactured by low temperature sintering technology[C]∥International Conference on Nuclear Engineering. US: American Society of Mechanical Engineers, 2013: V001T02A002.
[17]	GARNER F A, TOLOCZKO M B, MUNRO B, et al. Comparison of irradiation creep and swelling of an austenitic alloy irradiated in FFTF and PFR[C]∥Effects of Radiation on Materials: 18th International Symposium. US: ASTM International, 1999.
[18]	YVON P, CARR F. Structural materials challenges for advanced reactor systems[J]. Journal of Nuclear Materials, 2009, 385(2): 217-222.
[19]	GAFFARD V, BESSON J, GOURGUES-LORENZON A F. Creep failure model of a tempered martensitic stainless steel integrating multiple deformation and damage mechanisms[J]. International Journal of Fracture, 2005, 133(2): 139-166.
[20]	LUCUTA P G, HASTINGS I J. A pragmatic approach to modelling thermal conductivity of irradiated UO₂ fuel: Review and recommendations[J]. Journal of Nuclear Materials, 1996, 232(2-3): 166-180.
[21]	FINK J K. Thermophysical properties of uranium dioxide[J]. Journal of Nuclear Materials, 2000, 279(1): 1-18.
[22]	BALBOA H, VAN BRUTZEL L, CHARTIER A, et al. Assessment of empirical potential for MOX nuclear fuels and thermomechanical properties[J]. Journal of Nuclear Materials, 2017, 495: 67-77.
[23]	MONTGOMERY R, TOM C, LIU W, et al. Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis[J]. Journal of Computational Physics, 2017, 328: 278-300.
[24]	MIHAILA B, STAN M, RAMIREZ J, et al. Simulations of coupled heat transport, oxygen diffusion, and thermal expansion in UO₂ nuclear fuel elements[J]. Journal of Nuclear Materials, 2009, 394(2-3): 182-189.
[25]	van UFFELEN P, PASTORE G. Oxide fuel performance modeling and simulation[C]∥ KONINGS R. Comprehensive nuclear materials (Second Edition). Oak Ridge, TN: Elsevier, 2020: 363-416.
[26]	MILES K J. The SAS4A/SASSYS-1 safety analysis code system chapter 8: DEFORM-4: Steady-state and transient pre-failure pin behavior[R]. US: Argonne National Laboratory, 2017.
[27]	DEB K, PRATAP A, AGARWAL S, et al. A fast and elitist multiobjective genetic algorithm: NSGA-Ⅱ[J]. IEEE Transactions on Evolutionary Computation, 2002, 6(2):182-197.

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热管冷却反应堆固态堆芯热力耦合性能分析及结构优化

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Thermal-mechanical Coupling Analysis and Structure Optimization of Solid Core of Heat-pipe-cooled Reactor

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