基于统计抽样的压水堆燃料组件性能分析不确定度量化研究

为能在给出数值模拟结果的同时提供置信区间，本文开展了压水堆燃料性能分析、组件燃耗和热工水力学分析计算的不确定度量化研究。采用西安交通大学自主开发的不确定度分析程序平台NECP UNICORN，分别耦合了轻水堆燃料性能分析程序FEMAXI、压水堆群常数计算程序NECP Bamboo Lattice和热工水力子通道程序CTF。首先针对不同物理过程的特点，分析需要考虑的不确定度来源。然后针对核数据协方差矩阵稀疏且不满秩的特点，应用COST方法以减少样本量。结果表明，对于燃料性能分析，边界条件、几何参数和材料性质对燃料中心温度有显著影响。对于燃耗过程，核数据和几何参数对特征值、功率分布、两群常数和核子密度的不确定度有显著影响。对于热工水力分析过程，边界条件、几何参数和模型系数对冷却剂温度和包壳温度的不确定度有较大影响。针对每种物理场，分别量化其输入输出参数的不确定度，对于后续量化复杂系统多物理耦合过程的不确定度具有重要意义。

Uncertainty Quantification of PWR Fuel Assembly Performance Analysis Based on Statistical Sampling Method

The design and safe operation of nuclear reactor are inseparable from high-precision and high resolution simulation program. With the development of calculation methods and the progress of computer hardware, high fidelity numerical reactor technology becomes a research hotspot. However, due to the complexity of the objective world and the limitation of human cognitive level, there is uncertainty in any measurement and mathematical physical modeling process. Therefore, the simulation results need to provide the corresponding uncertainty as the confidence interval to meet the requirements of high confidence in numerical reactor technology. In order to promote the development of sensitivity and uncertainty analysis of reactor system, OECD/NEA (Organization for Economic Co operation and Development/Nuclear Energy Agency) issued the UAM LWRs (uncertainty analysis in modelling of light water reactors) benchmark. The exercises are carried out in 3 phases that is neutronics phase, core phase and system phase. In the neutronics phase, the uncertainty of nuclear data is transferred from the evaluation nuclear database files (ENDF) to multi group micro sections, and then to macro few group sections and key parameters of steady state core calculation. The core phase includes three independent physical processes which is fuel performance analysis, time dependent neutronics and bundle thermal hydraulic. The system phase includes multi physics coupling calculation, and finally gives the best estimation plus uncertainty (BEPU) result. In order to provide confidence interval while giving numerical simulation results, the uncertainty quantification of fuel performance, assembly burnup and thermal hydraulic analysis was carried out based on the second phase (core phase) of UAM LWRs benchmark. Using the uncertainty analysis platform NECP UNICORN independently developed by Xi’an Jiaotong University, the light water reactor fuel analysis code FEMAXI, the lattice physics program NECP Bamboo Lattice and the subchannel thermal hydraulic code CTF were coupled respectively. Firstly, according to the characteristics of different physical processes, the sources of uncertainty to be considered were analyzed. Then, according to the characteristics of sparse and unfilled rank covariance matrix of nuclear data, COST method was applied to reduce the sample size. It is found that for fuel performance analysis, boundary conditions, geometric parameters and material properties have significant effects on the uncertainty of fuel center temperature. For the burnup process, the nuclear data and manufacturing parameters have a significant impact on the uncertainty of eigenvalues, power distribution, two group constants and nucleon density. For thermal hydraulic analysis process, boundary conditions, geometric parameters and model coefficients have a great influence on the uncertainty of coolant and cladding temperature. For each physical field, quantifying the uncertainty of its input and output parameters is of great significance for quantifying the uncertainty of multi-physical coupling process of complex system.