Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT
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摘要: 针对各类小型动力堆或实验堆开展物理-热工耦合模拟计算时，由于非规则几何结构的存在而带来物理-热工网格映射关系复杂且不可统一预置的问题，基于数值反应堆高保真物理计算程序NECP-X开展了基于统一几何建模的物理-热工耦合方法研究，基于中子学模型建立物理-热工耦合的映射关系，并结合NECP-X程序中的瞬态计算方法实现了特殊功率偏移实验（SPERT）实验堆堆芯的直接瞬态计算；计算了SPERT实验堆稳态算例并与蒙特卡罗程序的结果进行对比，在此基础上，对SPERT实验堆进行了瞬态计算分析并与实验值进行对比。结果表明，NECP-X程序中子学计算的特征值和棒功率分布计算结果具有较高的精度；基于统一几何建模的网格映射方法可以方便快捷地实现复杂几何压水堆的物理-热工耦合计算；与实验值相比，瞬态计算的总功率、反应性随时间的变化曲线具有较高的精度，并且可提供精细的功率及温度分布。Abstract: To solve the problem that the thermal-neutronic grid mapping relation is complicated and cannot be preset in a centralized manner due to the existence of irregular geometry in performing the thermal-neutronic coupled simulation calculation for various small power reactors and experimental reactors, this paper studies the thermal-neutronic coupling method based on the unified geometric modeling, using the high-fidelity numerical code for reactor neutronics calculation, NECP-X. This study establishes the mapping relation for the thermal-neutronic coupling on the basis of the neutronics model, and enables the direct transient calculation of the experimental reactor core for the special power excursion reactor test (SPERT) via combination with the transient calculation method in NECP-X. Then, this study calculates the steady-state case for the experimental reactors of SPERT, and compares the calculation results with the results gained from the Monte Carlo code. On this basis, this study conducts transient calculation and analysis for these experimental reactors and compares the corresponding results with the experimental results. The final results show that the eigenvalues from the neutronics calculation by the NECP-X and the rod power distribution calculation results are of high accuracy; that the grid mapping method based on the unified geometric modeling allows a easy and fast thermal-neutronic coupled calculation of the PWRs of complex geometry; and that compared with the experimental values, the curve of change in the total power and reactivity gained from transient calculation with time is more accurate and can provide refined distributions of power and temperature.
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图 1 棒束型小堆中子学模型
1—燃料芯块；2—气隙区；3—包壳区；蓝色区域—慢化剂区
Figure 1. Neutronics Model of Rod-Cluster Reactor
图 3 NECP-X程序的空间区域分解示意图
红色部分—活性区；蓝色区域—轴向反射层区
Figure 3. Schematic Diagram of Space Division of NECP-X
图 4 NECP-X与并行热工程序耦合的分-总-分耦合策略
n、m—CPU个数
Figure 4. Coupling Strategy of NECP-X and Parallel Thermal Code
图 5 NECP-X与串行热工程序耦合分-总耦合策略
Figure 5. Coupling Strategy of NECP-X and Serial Thermal Code
图 7 SPERT实验堆堆芯径向几何模型
红色区域—燃料区；蓝色区域—慢化剂区；黄色区域—十字形瞬态棒；浅粉色区域——矩形控制棒；其他颜色区域为不同类型的结构材料
Figure 7. Radial Geometric Model of Experimental Reactor Core for SPERT
图 8 CZP工况SPERT实验堆三维全堆芯的棒功率及功率偏差分布
Figure 8. 3D Full-Core Pin Power and Power Difference Distributions of Experimental Reactor for SPERT in the CZP Case
图 9 HZP工况SPERT实验堆三维全堆芯的棒功率及功率偏差分布
Figure 9. 3D Full-Core Pin Power Distribution and Power Difference Distributions of Experimental Reactor for SPERT in the HZP Case
图 10 SPERT实验堆栅元类型图
红色区域—燃料区；粉色区域——矩形控制棒区；蓝色区域—慢化剂区；黄色区域—十字形瞬态棒区；其他颜色区域为不同类型的结构材料
Figure 10. Cell Types of Experimental Reactor for SPERT
图 13 活性区最底部初始以及峰值时刻燃料棒功率分布
Figure 13. Pin Power Distributions Corresponding to Initial and Peak Power at the Bottom of the Active Region
图 14 活性区最底部初始以及峰值时刻燃料棒温度分布
Figure 14. Pin Temperature Distributions Corresponding to Initital and Peak Temperature at the Bottom of the Active Region
表 1 SPERT实验堆三维全堆芯特征值计算结果
Table 1. 3D Full-Core Eigenvalue Calculation Results of Experimental Reactor for SPERT
工况 特征值 KENO-VI NECP-X 偏差/pcm CZP 0.99857±0.00001 0.99970 +113 HZP 1.00069±0.00001 1.00326 +257 
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表 2 T85算例工况状态参数
Table 2. Condition Parameters of T85 Case
反应性/$ 入口
温度/K入口质量
流速/(kg·s−1)初始
功率/MW最大
功率/MW0.87±0.04 533.15±2.22 598.6 19±1 130±10% 1 $= 6.605×10−3
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表 3 SPERT实验堆动力学参数计算结果
Table 3. Kinetic Parameters Calculations of Experimental Reactor for SPERT
参数 NECP-X 实验值 缓发中子份额 0.00730 0.00725 中子代时间/µs 17.2 16.31
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