确定论数值反应堆程序NECP-X的开发及应用

数值反应堆是基于大规模并行计算平台，利用先进的物理模型和数值模拟算法，采用精细化建模，从而精确模拟反应堆在正常运行与事故工况中发生的各类物理现象的模拟技术。西安交通大学NECP团队基于自研的多群和连续能量数据库，提出了全局 局部耦合输运计算方法、大规模并行的2D/1D耦合输运方法等，开发了基于确定论方法的数值反应堆物理程序NECP X，并在此基础上实现了物理 热工 燃料性能分析的多物理耦合模拟计算。基于该程序及其耦合系统，在商用大型压水堆、研究堆和实验堆中进行了验证应用。数值结果表明，NECP X程序及其耦合系统可准确预测反应堆在运行过程中的关键安全参数随时间的演变情况，如有效增殖因数、功率、温度、应力、间隙宽度等，可为商用大型压水堆、研究堆和研究堆的设计及安全分析提供可靠的工具。

Development and Application of Deterministic Numerical Reactor Code NECP-X

Based on the large-scale parallel computing platform, the numerical reactor uses advanced physical models, numerical simulation algorithms and fined modeling, so as to accurately simulate various physical phenomena of the reactor under normal operation and accident conditions. In order to solve the problems of low calculation resolution, low calculation accuracy and small application range caused by a series of theoretical approximations in the traditional reactor simulation technology, advanced simulation methods for numerical reactor were studied. In terms of neutronics resonance calculation, a resonance calculation method based on the global local coupling was proposed. The effects in resonance were divided into global effects and local effects. The Dancoff correction factor was used to deal with the global effects, and the pseudo resonant nuclide subgroup method was used to deal with the local effects. Based on the 2D/1D coupling method, an improved leakage splitting method was proposed to obtain the angular flux of each flat source region through the anisotropic angular reconstruction in the fuel rod, and a three level CMFD method was proposed to improve the computational efficiency of the traditional CMFD method. For the depletion calculations, a prediction/correction method based on reaction rate prediction was proposed to predict the change of nuclear reaction rate according to the first few burnup points and reduce the calculation time of subsequent burnup points. The Picard iterative method was used to calculate the full coupling of neutronics, thermal hydraulics and fuel performance. Based on the above methods, a numerical reactor code named NECP X was developed and verified by a set of internationally famous benchmark problems. It is proved that the code is of high calculation accuracy and efficiency. The code is then applied to the simulations of large commercial PWRs, research reactors and experimental reactors such as the second generation improved PWR M310, the third generation PWR AP1000, Xi’an Pulse Reactor, JRR-3M plate fuel reactor and Watts Bar reactor. For M310 core simulation, the critical boron concentration error of the code is within 30 ppm, and the maximum fuel effective temperature occurs at the beginning of the cycle and decreases with the burnup increasing. The change of the coolant temperature distribution is consistent with that of the fuel temperature distribution. Taking the solution from the Monte Carlo code as the reference, for the AP1000 simulation, the difference between NECP X and the reference in keff is 59 pcm. The difference in power for most assemblies is within 15%. For Xi’an Pulsed Reactor, the max difference in keff is 84 pcm, the max power difference is 52%. For JRR 3M plate reactor, the max difference in keff is 97 pcm, the max power difference is 18%. For the multiphysics coupled simulation of Watts Bar reactor, the max fuel temperature calculated by the code is 1 709 K, the Mises stress of cladding changes little with time and space, the gap width is large at both ends of the fuel rod and small in the middle. The numerical results show that it can accurately predict key safety parameters of the reactor core, such as effective multiplication factors, power distributions, temperatures, stress, gap width, etc., and it is a reliable tool for the design and safety analysis of commercial large scale pressurized water reactors, research reactors and research reactors.