Research on In-Pile Thermo-Mechanical Performance for U-10Mo/Zr Monolithic Fuel Element under Steady Condition
-
摘要: 本文建立了U-10Mo/Zr单片式燃料元件的辐照性能模型以及热-力学本构关系，采用有限元方法进行非均匀辐照场中燃料元件稳态热-力学性能的数值模拟，获得并分析了U-10Mo/Zr单片式燃料元件温度、形变和应力的分布特点及变化规律。研究结果表明，燃料芯体厚度增量在芯体和包壳结合面附近达到最大，主要受到燃料辐照蠕变的影响；在较低燃耗条件下，燃料芯体高温辐照肿胀模拟结果与低温辐照肿胀试验结果相当；燃料芯体边角区域和包壳端面外侧区域存在应力集中。Abstract: In the paper, the models of irradiation performance and thermo-mechanical constitutive relations of U-10Mo/Zr monolithic fuel element were established. With the finite element method, numerical simulation of fuel element thermo-mechanical performance under steady heterogeneous irradiation condition was conducted, the distribution and evolution characteristics of temperature, strain and stress in U-10Mo/Zr monolithic fuel element were acquired and analyzed. The results showed that the thickness increment of the fuel pellet becomes largest near the interface between fuel pellet and cladding, predominantly affected by fuel irradiation creep. Under the low burn-up conditions, the simulation result of fuel pellet swelling at high temperature condition equals the irradiation test results at low temperature. There are stress concentrations in the corner area of the fuel pellet and the outer area of the cladding end surface.
-
图 3 稳态运行80 d后燃料元件温度分布
Figure 3. Temperature Distribution of Fuel Element after 80 d of Steady-State Operation
图 4 燃料芯体在Z轴方向位移随运行时间分布
Figure 4. Distribution of Displacement of Fuel Pallet along Z Axis with Running Time
图 5 稳态运行80 d后燃料芯体应变量分布
Figure 5. Strain Distribution of Fuel Pallet after 80 d of Steady-state Operation
图 6 稳态运行80 d后燃料芯体ΔV/V分布
Figure 6. ΔV/V Distribution of Fuel Pallet after 80 d of Steady-state Operation
图 7 不同时刻燃料芯体最大ΔV/V值分布
Figure 7. Maximum ΔV/V Distribution of Fuel Pallet at Different Times
图 8 稳态运行80 d后燃料芯体边角附近包壳的Mises应力云图
Figure 8. Mises Stress Contour of Cladding around the Corner Area of the Fuel Pellet after 80 d of Steady-state Operation
图 9 稳态运行80 d后燃料芯体Mises应力云图
Figure 9. Mises Stress Contour of Fuel Pellet after 80 d of Steady-State Operation
表 1 燃料元件尺寸
Table 1. Size of Fuel Element
结构 长度/mm 宽度/mm 厚度/mm 燃料元件 420.0 70.0 1.5 燃料芯体 400.0 64.0 0.9 
下载: 导出CSV
-
[1] CRAWFORD D C, HAYES S L, POWERS J J. VTR startup fuel paper for NFSM: INL/EXT-18-44673-Rev000[R]. Idaho: Idaho National Laboratory, 2018. [2] DRAGUNOV Y G, TRETIYAKOV I T, LOPATKIN A V, et al. MBIR multipurpose fast reactor – Innovative tool for the development of nuclear power technologies[J]. Atomic Energy, 2012, 113(1): 24-28. doi: 10.1007/s10512-012-9590-x [3] KIM Y S, HOFMAN G L, CHEON J S, et al. Fission induced swelling and creep of U-Mo alloy fuel[J]. Journal of Nuclear Materials, 2013, 437(1-3): 37-46. doi: 10.1016/j.jnucmat.2013.01.346 [4] SHOUDY A A, MCHUGH W E, SILLIMAN M A. The effect of irradiation temperature and fission rate on the radiation stability of uranium-10 wt% molybdenum alloy[C]//Proceedings of Radiation Damage in Reactor Materials. Vienna: IAEA, 1963. [5] LÓPEZ M, PICCHETTI B, TABOADA H. Influence of temperature and compressive stress on the UMo/Zry-4 interdiffusion layer[J]. Progress in Nuclear Energy, 2017(94): 101-105. doi: 10.1016/j.pnucene.2016.10.006 [6] MILLER G K, BURKES D E, WACHS D M. Modeling thermal and stress behavior of the fuel-clad interface in monolithic fuel mini-plates[J]. Materials & Design, 2010, 31(7): 3234-3243. [7] YUN D, HOFMAN G L, KIM Y S, et al. Finite element modeling of irradiation induced swelling and creep in metallic mini-plate fuel - A preliminary study[J]. Transactions of the American Nuclear Society, 2011(105): 407-408. [8] ZHAO Y M, GONG X, DING S R. Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic U-Mo/Zr fuel plates under a heterogeneous irradiation condition[J]. Nuclear Engineering and Design, 2015(285): 85-97. [9] 殷明阳,庞华,唐昌兵,等. UMo-Zr单片式燃料板结构改进研究[J]. 核动力工程,2019, 40(4): 172-176. [10] 孔祥喆,丁淑蓉,田旭. UMo/Zr单片式燃料板堆内热力耦合行为研究[J]. 核动力工程,2018, 39(2): 109-113. [11] CUI Y, DING S R, CHEN Z T, et al. Modifications and applications of the mechanistic gaseous swelling model for UMo fuel[J]. Journal of Nuclear Materials, 2015(457): 157-164. doi: 10.1016/j.jnucmat.2014.11.065 [12] WILLARD R M, SCHMITT A R. Irradiation swelling, phase reversion, and intergranular cracking of U-10wt. % Mo fuel alloy: NAA-SR-8956[R]. California: Atomics International, 1964. [13] BLEIBERG M L. Effect of fission rate and lamella spacing upon the irradiation-induced phase transformation of U-9wt% Mo alloy[J]. Journal of Nuclear Materials, 1959, 1(2): 182-190. doi: 10.1016/0022-3115(59)90051-0 [14] KIM Y S, HOFMAN G L. Fission product induced swelling of U-Mo alloy fuel[J]. Journal of Nuclear Materials, 2011, 419(1-3): 291-301. doi: 10.1016/j.jnucmat.2011.08.018 [15] YAN F, JIAN X B, DING S R. Effects of UMo irradiation creep on the thermo-mechanical behavior in monolithic UMo/Al fuel plates[J]. Journal of Nuclear Materials, 2019(524): 209-217. doi: 10.1016/j.jnucmat.2019.07.006 [16] FISHER F E, RENKEN J C. Single-crystal elastic moduli and the Hcp→bcc transformation in Ti, Zr, and Hf[J]. Physical Review A, 1964, 135(2A): A482-A494. doi: 10.1103/PhysRev.135.A482 [17] HAGRMAN D L, REYMAN G A. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior: NUREG/CR-0497[R]. Idaho: Idaho National Engineering Laboratory, 1979. [18] HALES J D, WILLIAMSON R L, NOVASCONE S R, et al. BISON theory manual the equations behind nuclear fuel analysis: INL/EXT-13-29930[R]. Idaho: Idaho National Laboratory, 2016. [19] JAEGER W. Heat transfer to liquid metals with empirical models for turbulent forced convection in various geometries[J]. Nuclear Engineering and Design, 2017(319): 12-27. doi: 10.1016/j.nucengdes.2017.04.028 [20] REST J, KIM Y S, HOFMAN G L, et al. U-Mo Fuels Handbook: ANL-09/31[R]. Argonne: Argonne National Laboratory, 2006. -
计量
- 文章访问数: 316
- HTML全文浏览量: 119
- PDF下载量: 72
- 被引次数: 0
