束流瞬变下CiADS次临界反应堆燃料包壳响应特性研究
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摘要
通过修改系统分析程序RELAP5 MOD4.0的点堆动力学模型与流动传热模型,使其具备了模拟液态铅铋冷却次临界反应堆动力学特性的能力;利用改进的程序模拟了加速器驱动嬗变研究装置(CiADS)的次临界反应堆燃料包壳在发生束流瞬变时的响应特性;利用ANSYS17.0程序分析了CiADS次临界反应堆燃料包壳束流瞬变下的应力变化。研究表明:失束时间越短,燃料包壳的温度回升越慢;燃料包壳不会因可能发生的束流超功率事件而发生熔毁;燃料包壳内外壁面的温差变化是影响应力变化的主要因素;CiADS次临界反应堆的燃料包壳不会因束流瞬变而发生应力破坏。
The point kinetic model and the heat transfer model of the reactor system analysis code RELAP5 MOD4.0 were modified,which provided the capacity to simulate the dynamic characteristics of the liquid lead-bismuth cooled sub-critical reactor.The response characteristics of the fuel rods in the sub-critical reactor of China Initiative Accelerator Driven System(CiADS)under beam transient were simulated by the improved code RELAP5 MOD4.0.The variation of stress of the fuel cladding in CiADS under beam transient was analyzed by ANSYS17.0 code.The results show that the shorter the time of beam trip the slower the temperature of the fuel rod recovery.The fuel cladding will not melt in the possible overpower event of the beam.The change of temperature difference between inner wall and outer wall of fuel cladding is the main factor of stress variation.The beam transient will not cause stress damage to the fuelcladding in CiADS sub-critical reactor.
The point kinetic model and the heat transfer model of the reactor system analysis code RELAP5 MOD4.0 were modified,which provided the capacity to simulate the dynamic characteristics of the liquid lead-bismuth cooled sub-critical reactor.The response characteristics of the fuel rods in the sub-critical reactor of China Initiative Accelerator Driven System(CiADS)under beam transient were simulated by the improved code RELAP5 MOD4.0.The variation of stress of the fuel cladding in CiADS under beam transient was analyzed by ANSYS17.0 code.The results show that the shorter the time of beam trip the slower the temperature of the fuel rod recovery.The fuel cladding will not melt in the possible overpower event of the beam.The change of temperature difference between inner wall and outer wall of fuel cladding is the main factor of stress variation.The beam transient will not cause stress damage to the fuelcladding in CiADS sub-critical reactor.
引文
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[3]彭天骥,顾龙,王大伟,等.中国加速器驱动嬗变研究装置次临界反应堆概念设计[J].原子能科学技术,2017,51(12):2 235-2 241.PENG Tianji,GU Long,WANG Dawei,et al.Conceptual design of subcritical reactor for China initiative accelerator driven system[J].Atomic Energy Science and Technology,2017,51(12):2 235-2 241(in Chinese).
[4]ABDERRAHIM H A,GALAMBOS J,GOHAR Y,et al.Accelerator and target technology for accelerator driven transmutation and energy production[M/OL].2010[2017-08-20].http:∥pdfs.semanticscholar.org/55e9/b442e7b189a4f-31fefa6939ce3387d0c47b7.pdf.
[5]LIU P,CHEN X,RINEISKI A,et al.Transient analyses of the 400 MWth-class EFIT accelerator driven transmuter with the multi-physics code:SIMMER-Ⅲ[J].Nuclear Engineering and Design,2010,240(10):3 481-3 494.
[6]SUGAWARA T,NISHIHARA K,TSUJIMOTO K.Transient analysis for lead-bismuth cooled accelerator-driven system[J].Annals of Nuclear Energy,2013,55:238-247.
[7]AHMAD A,LINDLEY B A,PARKS G T.Accelerator-induced transients in accelerator driven subcritical reactors[J].Nuclear Instrument and Methods in Physics Research,2012,696(6):55-65.
[8]汪振,王刚,辜峙钘,等.加速器驱动铅铋冷却自然循环次临界堆束流瞬变事故研究[J].核技术,2015,38(1):91-96.WANG Zhen,WANG Gang,GU Zhixing,et al.Beam transient accident for lead-bismuth cooled ADS by natural circulation[J].Nuclear Technique,2015,38(1):91-96(in Chinese).
[9]XUE S,JIN M,WANG G,et al.Fuel cladding integrity analysis during beam trip transients for China lead-based demonstration reactor[J].Annals of Nuclear Energy,2015,83:94-100.
[10]谢仲生,吴宏春,张少泓.核反应堆物理分析[M].西安:西安交通大学出版社,2004:228-251.
[11]蒋校丰.加速器驱动次临界系统中子学计算方法的研究及快-热耦合系统的概念设计[D].西安:西安交通大学,2005.
[12]CHEN S K,TODREAS N.Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles:Bundle friction factors[J].Nuclear Engineering and Design,1986,92:227-251.
[13]吕科峰.液态铅铋合金在带绕丝棒束组件内热工水力行为研究[D].合肥:中国科学技术大学,2016.
[14]BORISHANSKII V M,GOTOVSKII M A,FIRSOVAV.Heat transfer to liquid metals in longitudinally wetted bundles of rods[J].Soviet Atomic Energy,1969,27(6):1 347-1 350.
[15]FONT E B.IFMIF accelerator facility RAMI analyses in the engineering design phase[D].Spain:Universitat Politècnica de Catalunya,2014.
[16]AFCEN.RCC-MR[S].France:AFCEN,2007.
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