Study on Simulation Technology of Cracking and Blistering of Particle Dispersed Plate Fuel Element
-
摘要: 颗粒弥散板状燃料元件在堆内运行时，可能发生开裂起泡现象，影响燃料及核反应堆的安全性。针对颗粒弥散板状燃料元件进行了开裂起泡现象的数值模拟研究。为了实现从细观燃料颗粒到宏观燃料元件的开裂分析和起泡模拟，首先建立了一套多维度跨尺度耦合的模拟分析策略，使用从细观燃料颗粒尺度到宏观元件尺度的跨尺度耦合方案，实现了从燃料颗粒出发的燃料元件芯体开裂危险区域判定；基于包含了三维燃料元件、一维冷却剂和二维燃料切片的多维度耦合体系，应用扩展有限元方法，实现了燃料元件的二维开裂起泡模拟；为了准确判断起泡的发生，建立了一套裂变气体内压模型，与二维扩展有限元相结合，最终实现了热冲击工况下的宏观燃料元件的裂纹扩展模拟，并预测当燃耗为120 MW·d·kg−1(U)时，起泡发生的温度为790 K左右。Abstract: Cracking and blistering may occur when the particle dispersed plate fuel element runs in the reactor, which affects the safety of fuel and nuclear reactor. In this paper, numerical simulation research on cracking and blistering of particle dispersed plate fuel element is carried out. To realize the cracking analysis and blistering simulation from the fine-scale fuel particles to the macro-scale fuel elements, this paper first establishes a set of multi-dimensional multi-scale coupled simulation methods. With the multi-scale coupling scheme from fine-scale fuel particles to the macro-scale fuel element, the fuel element cracking risk region determination is realized. Based on the multi-dimensional coupling system that includes a 3D fuel element, a 1D coolant, and a 2D fuel slice, the extended finite element method (XFEM) is applied to realize the 2D cracking and blistering simulation of the fuel element. To accurately judge the occurrence of blistering, a fission gas internal pressure model is established in this paper, which is combined with the 2D extended finite element. Finally, the crack extension simulation of macroscopic fuel elements under thermal shock conditions is realized, and the temperature at which blistering occurs is predicted to be about 790 K when the fuel burnup is 120 MW·d·kg−1(U).
-
图 1 细观颗粒-宏观元件跨尺度耦合开裂判定方法
Figure 1. Multi-scale Coupling from Fine-scale Particle to Macro-scale Element for Cracking Determination Method
图 2 三维燃料元件-一维冷却剂-二维燃料切片多维度耦合
Figure 2. Multi-Dimensional Coupling of 3D Fuel Element-1D Coolant-2D Fuel Slice
图 6 燃料元件起泡过程裂变气体内压变化
Figure 6. Variation of Fission Gas Internal Pressure during Fuel Element Blistering
图 7 含随机分布气孔的燃料颗粒几何模型及网格
Figure 7. Geometric Model and Meshing of Fuel Particle with Randomly Distributed Pores
图 8 含随机分布气孔的燃料颗粒vonMises应力分布
Figure 8. VonMises Stress Distribution in Fuel Particle with Randomly Distributed Pores
图 9 颗粒vonMises应力峰值-温度-燃耗关系
Figure 9. VonMises Peak Stress-Temperature-Burnup Relationship for Fuel Particles
图 10 基于宏观条件的燃料颗粒开裂判定
Figure 10. Determination of Fuel Particle Cracking Based on Macroscopic Conditions
图 11 燃料元件内部颗粒开裂危险区域演化
Figure 11. Evolution of Particle Cracking Risk Region in Fuel Element
图 12 燃料元件多维度耦合开裂起泡模拟
Figure 12. Simulation of Fuel Element Cracking and Blistering with Multi-dimensional Coupling
图 14 燃料元件热冲击瞬态工况局部起泡模拟结果
Figure 14. Simulation Results of Localized Blistering of Fuel Elements in Thermal Shock Transient
图 15 燃料起泡过程裂变气体内压与热冲击温度关系
Figure 15. Curve of Fission Gas Internal Pressure Versus Temperature for Fuel Blistering in Thermal Shock Transient
图 16 燃料元件起泡高度与热冲击温度关系曲线
Figure 16. Curve of Fuel Blistering Height Versus Temperature in Thermal Shock Transient
图 17 燃料起泡直径与热冲击温度关系曲线
Figure 17. Curve of Fuel Blistering Diameter Versus Temperature in Thermal Shock Transient
-
[1] 高利军,陈炳德,姜胜耀,等. 弥散型燃料板的辐照起泡机理分析[J]. 原子能科学技术,2012, 46(S2): 819-825. [2] 龙冲生,赵毅,高雯,等. 基于断裂强度的陶瓷燃料颗粒开裂模型[J]. 核动力工程,2014, 35(1): 92-96,105. [3] 赵毅,龙冲生,王晓敏. 弥散燃料颗粒裂纹起源的有限元模拟分析[J]. 原子能科学技术,2015, 49(2): 311-315. [4] 陈洪生,龙冲生,肖红星,等. 裂变气体气泡尺寸对弥散燃料颗粒内部特征的影响规律[J]. 核动力工程,2018, 39(2): 27-31. [5] 陈洪生,龙冲生,肖红星,等. 基于弥散燃料颗粒开裂的裂变气体释放模型[J]. 核动力工程,2019, 40(5): 85-91. [6] 陈洪生,龙冲生,肖红星. 基于弥散燃料颗粒开裂的金属基体裂纹特征模型[J]. 原子能科学技术,2020, 54(2): 334-339. [7] 严峰,丁淑蓉,李垣明,等. UMo/Zr单片式燃料板起泡行为数值模拟[J]. 原子能科学技术,2018, 52(6): 1063-1069. [8] GEELHOOD K J, LUSCHER W G, RAYNAUD P A, et al. FRAPCON-4.0 - a computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup: PNNL-19418, Vol.1 Rev.2[R]. Washington: U. S. Department of Energy, 2015. [9] GEELHOOD K J, LUSCHER W G. FRAPTRAN-1.5 - integral assessment: PNNL-19400, Vol. 2, Rev. 1[R]. Richland: Pacific Northwest National Laboratory, 2014. [10] DENG Y B, WU Y W, LI Y M, et al. Mechanism study and theoretical simulation on heat split phenomenon in dual-cooled annular fuel element[J]. Annals of Nuclear Energy, 2016, 94: 44-54. doi: 10.1016/j.anucene.2016.02.019 [11] DENG Y B, WU Y W, ZHANG D L, et al. Thermal-mechanical coupling behavior analysis on metal-matrix dispersed plate-type fuel[J]. Progress in Nuclear Energy, 2017, 95: 8-22. doi: 10.1016/j.pnucene.2016.11.007 [12] DENG Y B, WU Y W, GONG C, et al. Upgrade of FROBA code and its application in thermal-mechanical analysis of space reactor fuel[J]. Nuclear Engineering and Design, 2018, 332: 297-306. doi: 10.1016/j.nucengdes.2018.03.041 [13] HALES J D, NOVASCONE S R, SPENCER B W, et al. Verification of the BISON fuel performance code[J]. Annals of Nuclear Energy, 2014, 71: 81-90. doi: 10.1016/j.anucene.2014.03.027 [14] WILLIAMSON R L, CAPPS N A, LIU W, et al. Multi-dimensional simulation of LWR fuel behavior in the BISON fuel performance code[J]. JOM, 2016, 68(11): 2930-2937. doi: 10.1007/s11837-016-2115-7 [15] HE Y A, CHEN P, WU Y W, et al. Preliminary evaluation of U3Si2-FeCrAl fuel performance in light water reactors through a multi-physics coupled way[J]. Nuclear Engineering and Design, 2018, 328: 27-35. [16] 邓超群,向烽瑞,贺亚男,等. 基于MOOSE平台的棒状燃料元件性能分析程序开发与验证[J]. 原子能科学技术,2021, 55(7): 1296-1303. [17] 邓超群,向烽瑞,贺亚男,等. 基于MOOSE平台的棒状燃料元件性能瞬态分析程序开发与验证[J]. 原子能科学技术,2021, 55(8): 1429-1439. [18] HE Y A, NIU Y H, XIANG F R, et al. Preliminary development of a multi-physics coupled fuel performance code for annular fuel analysis under normal conditions[J]. Nuclear Engineering and Design, 2022, 393: 111810. [19] XIANG F R, HE Y A, NIU Y H, et al. A new method to simulate dispersion plate-type fuel assembly in a multi-physics coupled way[J]. Annals of Nuclear Energy, 2022, 166: 108734. [20] XIANG F R, HE Y A, WU Y W, et al. Investigation of plate fuel performance under reactivity initiated accidents with developed multi-dimensional coupled method[J]. Journal of Nuclear Materials, 2023, 583: 154537. [21] LIU R, PRUDIL A, ZHOU W Z, et al. Multiphysics coupled modeling of light water reactor fuel performance[J]. Progress in Nuclear Energy, 2016, 91: 38-48. doi: 10.1016/j.pnucene.2016.03.030 [22] WANG Y Y, GUO Y H, WU Y W, et al. Preliminary analysis on the thermal-mechanical behavior of dispersed plate-type fuel under reactivity insertion accident[J]. Annals of Nuclear Energy, 2021, 163: 108509. doi: 10.1016/j.anucene.2021.108509 [23] NASIR R, MIRZA N M, MIRZA S M. Sensitivity of reactivity insertion limits with respect to safety parameters in a typical MTR[J]. Annals of Nuclear Energy, 1999, 26(17): 1517-1535. doi: 10.1016/S0306-4549(99)00038-9 [24] MIRZA A M, KHANAM S, MIRZA N M. Simulation of reactivity transients in current MTRs[J]. Annals of Nuclear Energy, 1998, 25(18): 1465-1484. doi: 10.1016/S0306-4549(98)00020-6 [25] KHATER H, ABU-EL-MATY T, EL-DIN EL-MORSHDY S. Thermal-hydraulic modeling of reactivity accidents in MTR reactors[J]. Nuclear Technology and Radiation Protection, 2006, 21(2): 21-32. doi: 10.2298/NTRP0602021K [26] IAEA. Research reactor core conversion from the use of highly enriched uranium fuels: guidebook: IAEA-TECDOC-233[R]. Vienna: International Atomic Energy Agency, 1980. [27] IAEA. Research reactor core conversion guidebook: volume 4: fuels (appendices I-K): IAEA-TECDOC-643[R]. Vienna: International Atomic Energy Agency, 1992. [28] LUCUTA P G, MATZKE H J, 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. doi: 10.1016/S0022-3115(96)00404-7 [29] HARDING J H, MARTIN D G. A recommendation for the thermal conductivity of UO2[J]. Journal of Nuclear Materials, 1989, 166(3): 223-226. doi: 10.1016/0022-3115(89)90218-3 [30] SIEFKEN L J, CORYELL E W, HARVEGO E A, et al. SCDAP/RELAP5/MOD 3.3 code manual: MATPRO-A library of materials properties for light-water-reactor accident analysis: Technical Report NUREG/CR-6150[R]. Washington: Division of Systems Technology, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, 2001. [31] ALLISON C M, BERNA G A, CHAMBERS R, et al. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO–A library of materials properties for light-water-reactor accident analysis: NUREG/CR-6150[R]. Washington: Idaho National Engineering Laboratory, 1993. [32] MAXWELL J C. A treatise on electricity and magnetism[M]. Oxford: Clarendon Press, 1873: 365. [33] 姜馨. 弥散型燃料的等效性质及棒状元件的辐照力学行为的研究[D]. 上海: 复旦大学,2009. [34] 杨烁. 核燃料元件内陶瓷颗粒的开裂行为模拟及PCI失效行为研究[D]. 西安: 西安交通大学,2021. [35] JIANG W, SPENCER B W, DOLBOW J E. Ceramic nuclear fuel fracture modeling with the extended finite element method[J]. Engineering Fracture Mechanics, 2020, 223: 106713. doi: 10.1016/j.engfracmech.2019.106713 [36] 伍晓勇,王斐,温榜. UO2弥散型燃料辐照后高温失效时显微分析[J]. 核动力工程,2012, 33(1): 74-77. -
下载:
计量
- 文章访问数: 38
- HTML全文浏览量: 15
- PDF下载量: 16
- 被引次数: 0
