Verification of Sodium-cooled Fast Reactor SUPERFACT-1 SF4/SF16 Fuel Rod Experiment using LoongCALF Code
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摘要: 由于钠冷快堆功率密度高、燃耗深度大，其燃料在运行过程中具有温度高、裂变气体释放率高、形变大、形成中心空洞等特点，因此钠冷快堆对燃料性能程序开发提出了新的挑战。LoongCALF程序是基于有限元方法和JFNK方法的快堆燃料性能分析程序，为验证LoongCALF程序在钠冷快堆燃料性能分析中模型的适用性，本文运用LoongCALF程序对SUPERFACT-1辐照实验中SF4/SF16燃料棒进行模拟，并将模拟结果与公开文献中TRANSURNUS、GERMINAL、MACROS等快堆燃料性能程序的结果对比。研究结果表明，LoongCALF程序计算得出的包壳温度、燃料棒内压以及芯块温度与文献结果符合较好，轴向中心空洞直径与实验结果符合较好，能够满足对钠冷快堆模拟的需求。因此，LoongCALF程序能够用于钠冷快堆的模拟工作，但裂变气体释放与气隙宽度以及核素分布等相关模型还需要进一步完善。
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关键词:
- LoongCALF程序 /
- 燃料性能分析 /
- 钠冷快堆 /
- 验证分析
Abstract: Due to the high power density and deep burn-up of sodium-cooled fast reactor, its fuel has the characteristics such as high temperature, high fission gas release rate, large deformation, and the formation of central voids during operation. Therefore, the sodium-cooled fast reactor poses new challenges to the development of fuel performance codes. The LoongCALF code is a fast reactor fuel performance analysis code based on the finite element method and JFNK method. To verify the applicability of the LoongCALF code in the analysis of sodium-cooled fast reactor fuel performance, this work uses the LoongCALF code to simulate the SF4/SF16 fuel rods in the SUPERFACT-1 irradiation experiment, and compares the simulation results with those of fast reactor fuel performance codes such as TRANSURNUS, GERMINAL and MACROS in public literature. The research results show that the cladding temperature, fuel rod internal pressure, and pellet temperature calculated by the LoongCALF code are in good agreement with the literature results, and the axial central void diameter is in good agreement with the experimental results, which can meet the needs of sodium-cooled fast reactor simulation. Therefore, the LoongCALF code can be used for the simulation work of sodium-cooled fast reactors, but the related models of fission gas release and gap width need to be further improved.-
Key words:
- LoongCALF code /
- Fuel performance analysis /
- Sodium-cooled fast reactor /
- Verification
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图 3 峰值功率节点上的燃料中心线温度演变
Figure 3. Fuel Centerline Temperature Evolution at the Peak Power Node
图 5 寿期末燃料轴向中心空洞直径分布
Figure 5. Diameter Distribution of Axial Central Void of Fuel at the End of Life
图 8 峰值功率节点上的燃料棒内压演变
Figure 8. Fuel Rod Internal Pressure Evolution at the Peak Power Node
图 9 满功率运行1 h后的轴向包壳温度分布
Figure 9. Axial Profiles of Cladding Temperature after 1 h Operation under Full Power
表 1 SUPERFACT-1实验考虑的燃料棒的设计参数
Table 1. Design Parameters of Fuel Rods Considered in SUPERFACT-1 Experiment
参数名 SF7/SF13 SF4/SF16 芯块半径/mm 2.68 2.71 径向间隙/mm 0.143 0.116 芯块理论密度/% 97.5 96.3 Pu与金属的质量比（Pu/M） 0.244 0.237 O与金属的原子比（O/M） 1.943 1.957 燃料长度/mm 850 包壳材料 15-15Ti冷作奥氏体不锈钢 包壳厚度/mm 0.45 上气腔体积/mm³ 1930 下气腔体积/mm³ 19530 氦气压力/MPa 0.1 氦气温度/℃ 20 
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表 2 钠冷却剂性能的详细参数
Table 2. Detailed Parameters of Sodium Coolant Performance
参数名 参数值 钠质量流量/(kg·s−1) 0.098 钠进口温度/℃ 395 钠压力/MPa 0.1
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表 3 SUPERFACT-1燃料棒的线功率轴向节点化
Table 3. Axial Nodalization Applied to Line Power of Fuel Rods in SUPERFACT-1
轴向节数 节点高度/mm 平均线功率因子 1 42.5 0.7396 2 127.5 0.9494 3 212.5 1.1182 4 297.5 1.2341 5 382.5 1.2882 6 467.5 1.2663 7 552.5 1.1749 8 637.5 1.0331 9 722.5 0.8477 10 807.5 0.6415
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表 4 Polypole算法计算参数
Table 4. Calculation Parameters of Polypole Algorithm
参数名 设定值 算法的收敛限值 0.01 算法在每个时间步的采样点个数 1 算法的最大迭代次数 5 裂变所产生气泡覆盖晶界表面的比例 0.5 算法使用的展开函数的阶数 5
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表 5 Picard模拟计算的计算参数
Table 5. Calculation Parameters of Picard Simulation Calculations
参数名 设定值 非线性热学求解的绝对误差 10−5 非线性力学求解的绝对误差 10−5 线性热学求解的绝对误差 10−5 线性力学求解的绝对误差 10−5 热学求解的最大线性迭代次数 50 力学求解的最大线性迭代次数 50 热学求解的最大非线性迭代次数 50 力学求解的最大非线性迭代次数 50 Picard求解中燃料最大温度的误差 0.05 最大Picard迭代次数 50
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[1] VAN UFFELEN P, SCHUBERT A, VAN DE LAAR J, et al. Verification of the TRANSURANUS fuel performance code—an overview[C]//Proceedings of 7th International Conference on WWER Fuel Performance, Modelling and Experimental Support. Albena, Bulgaria: INIS, 2007: 305. [2] TSUBOI Y, ENDO H, ISHIZU T, et al. Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code ‘FEMAXI-FBR’[J]. Journal of Nuclear Science and Technology, 2012, 49(4): 408-424. doi: 10.1080/00223131.2012.669242 [3] MELIS J C, ROCHE L, PIRON J P, et al. GERMINAL—A computer code for predicting fuel pin behaviour[J]. Journal of Nuclear Materials, 1992, 188: 303-307. doi: 10.1016/0022-3115(92)90488-7 [4] 黄晨. 快堆燃料棒寿命分析程序[D]. 北京: 中国原子能科学研究院,2000. [5] 齐飞鹏. 液态金属冷却快堆燃料元件的性能分析程序开发及应用[D]. 合肥: 中国科学技术大学,2018. [6] 许晓北,刘宙宇,曹良志,等. 基于MOOSE平台的燃料性能分析程序开发[J]. 核动力工程,2020, 41(S1): 130-134. [7] 吴宏春,杨红义,曹良志,等. 金属冷却快堆关键分析软件的现状与展望[J]. 现代应用物理,2021, 12(1): 010201. [8] LUZZI L, BARANI T, BOER B, et al. Assessment of three European fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment[J]. Nuclear Engineering and Technology, 2021, 53(10): 3367-3378. doi: 10.1016/j.net.2021.04.010 [9] 邵世豪,刘宙宇,许晓北,等. 基于MOOSE平台液态金属冷却快堆燃料性能分析程序开发[J]. 原子能科学技术,2024, 58(3): 539-548. doi: 10.7538/yzk.2024.youxian.0001 -
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