HCF Three-dimensional Refined Burnup Characteristics Analysis
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摘要: 几何形状复杂的螺旋十字燃料(HCF)对燃耗特性研究提出了更高的挑战。传统的同心圆圈式燃耗区域划分方法无法准确地模拟HCF复杂的几何引起的燃料各位置燃耗不同的问题,缺乏相应的三维精细化数值分析方法预测燃耗特性。本文针对HCF提出六面体燃耗区域划分及计算机辅助设计(CAD)几何建模方法,分别以HCF的薄片、最小扭转单元、单根燃料为研究对象,实现三维精细化的燃耗计算,获得不同燃耗下变量分布及235U、 238U、239Pu典型核素在凹、凸处的核子密度及核反应率。结果表明, HCF径向一周快中子通量密度、热中子通量密度、功率密度分布表现出极大的不均匀性,且随着燃料的消耗,其周向不均匀性增强,凸处燃耗较凹处深15.92 MW·d/kg。轴向扭转对燃料凸处物理变量的影响大于凹处。三维精细化的燃耗特性分析可为高保真的HCF中子物理和热工水力、力学等耦合计算提供基础。
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关键词:
- 螺旋十字燃料(HCF) /
- 燃耗计算 /
- 燃耗区域划分 /
- 功率密度
Abstract: Helical cruciform fuel (HCF) has a complex geometry, which poses a higher challenge to the study of burnup characteristics. The traditional concentric circle burnup region division method could not accurately simulate the different burnup at different positions caused by the complex geometry of HCF in the burnup, and lack corresponding three-dimensional refined numerical analysis method to predict the burnup characteristics. In this paper, a hexahedral burnup region division and Computer-Aided Design (CAD) geometric modeling method are proposed for HCF. By taking the slice, minimum twist unit and single fuel of HCF as the research object, three-dimensional burnup calculation are realized, and variable distribution under different burnups, nuclear density and nuclear reaction rate of typical nuclides 235U, 238U and 239Pu at concave and convex position are obtained. The results show that the fast neutron flux, thermal neutron flux and power density distribution of radial circumferential HCF show great non-uniformity. The circumferential non-uniformity increases with the depletion of fuel, the burnup in the convex position is 15.92 MW·d/kg deeper than that in the concave. The influence of axial twist on the physical variables of the convex position of fuel is greater than that of the concave position. Three-dimensional refined analysis of burnup characteristics provides a basis for high-fidelity coupling calculation of neutronic physics, thermal-hydraulics and mechanics of HCF.-
Key words:
- HCF /
- Burnup calculation /
- Burnup region division /
- Power density
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表 1 燃料元件各区域核素成分及材料密度
Table 1. Nuclide Composition and Material Density in Each Region of Fuel Element
区域 核素 密度/(g·cm−3) 方形毒物区域 90Zr, 91Zr, 92Zr, 94Zr, 96Zr,157Gd,93Nb 6.5075 燃料棒 90Zr, 91Zr, 92Zr, 94Zr,96Zr,235U,238U,241Am,228Th,229Th,
239Th,232Th,235Np,236Np,237Np,252Cf9.7012 包壳 90Zr, 91Zr, 92Zr, 94Zr, 96Zr,118Sn 6.4987 冷却剂 1H,16O 0.7172 表 2 不同燃耗步HCF薄片最外侧周向平均功率密度q
Table 2. Outer Circumferential Average Power Density Distribution for the HCF Slice in Different Burnup Steps
燃耗时间 凸处q/(W·cm−3) 凹处q/(W·cm−3) 0 h 827 734 6 h 815 726 36 h 821 728 100 d 36 h 828 732 700 d 36 h 881 758 -
[1] SHAH M D, GODDARD B, LLOYD C, et al. Investigating material attractiveness for an innovative metallic fuel design[J]. Nuclear Engineering and Design, 2020, 357: 110385. doi: 10.1016/j.nucengdes.2019.110385 [2] HARDY M M. Geometric transformation for double helical wire rods[D]. Oahu: University of Hawaii, 2004. [3] BRITT T, GODDARD B, SHAH M. Innovative fuel design to improve proliferation resistance[J]. Journal of Nuclear Materials Management, 2019, 47(3): 5-8. [4] 谢仲生. 核反应堆物理分析[M]. 北京: 原子能出版社,1981: 169-190. [5] SHIRVAN K, KAZIMI M S. Nuclear design of helical cruciform fuel rods[C]//Conference on Advances in Reactor Physics - Linking Research, Industry, and Education. Knoxville: American Nuclear Society, Inc. , 2012. [6] SILVA R H M, DE SOUZA W F, DA SILVA C A M, et al. Neutronic and thermal hydraulic analysis of a PWR fuel assembly using uranium mononitride[J]. Annals of Nuclear Energy, 2023, 194: 110071. doi: 10.1016/j.anucene.2023.110071 [7] AKTER Y, SAHADATH M H, REZA F. Assessment of the burnup characteristics of UO2 and MOX fuel in the mixed solid and annular rod configuration[J]. Nuclear Engineering and Design, 2021, 381: 111339. doi: 10.1016/j.nucengdes.2021.111339 [8] SAENZ B L, JEVREMOVIC T. Burn-cycle-dependent spatial distribution of nuclear fuel actinides and fission products based on the AGENT code[J]. Annals of Nuclear Energy, 2021, 160: 108390. doi: 10.1016/j.anucene.2021.108390 [9] KIM K M, CHOI N, LEE H G, et al. Practical methods for GPU-based whole-core Monte Carlo depletion calculation[J]. Nuclear Engineering and Technology, 2023, 55(7): 2516-2533. doi: 10.1016/j.net.2023.04.021 [10] CLARNO K T, PHILIP B, COCHRAN W K, et al. The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code[J]. Nuclear Engineering and Design, 2012, 252: 108-120. doi: 10.1016/j.nucengdes.2012.07.018 [11] EBIWONJUMI B, LEE H, KIM W, et al. Validation of nuclide depletion capabilities in Monte Carlo code MCS[J]. Nuclear Engineering and Technology, 2020, 52(9): 1907-1916. doi: 10.1016/j.net.2020.02.017 [12] ISOTALO A, SAHLBERG V. Comparison of neutronics-depletion coupling schemes for burnup calculations[J]. Nuclear Science and Engineering, 2015, 179(4): 434-459. doi: 10.13182/NSE14-35 [13] GOORLEY T, JAMES M, BOOTH T, et al. Initial MCNP6 release overview[J]. Nuclear Technology, 2012, 180(3): 298-315. doi: 10.13182/NT11-135 [14] ROMANO P K, HORELIK N E, HERMAN B R, et al. OpenMC: a state-of-the-art Monte Carlo code for research and development[J]. Annals of Nuclear Energy, 2015, 82: 90-97. doi: 10.1016/j.anucene.2014.07.048 [15] LEPPÄNEN J, PUSA M, VIITANEN T, et al. The Serpent Monte Carlo code: status, development and applications in 2013[J]. Annals of Nuclear Energy, 2015, 82: 142-150. [16] ZHENG M Y, TIAN W X, WEI H Y, et al. Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment[J]. Annals of Nuclear Energy, 2014, 63: 491-498. doi: 10.1016/j.anucene.2013.08.020 -