使用MCNP和COMSOL多物理场测量反应堆级石墨中热中子扩散长度的报告
Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics
摘要：
从实验和理论两方面测量了中子在反应堆级石墨中的扩散长度。实验工作包括蒙特卡罗（MC）的编码使用'MCNP'和有限元分析（FEA）的编码使用'COMSOL Multiphysics'和Matlab。采用MCNP程序模拟了2m × 2m微浓缩铀慢化堆（$^{235}U$）中的热中子扩散长度，建立了基于偏微分方程和常微分方程的点动力学模型。理论工作包括数值逼近方法，包括超越技术来说明迭代过程与有限元方法。最后，测量了热中子在石墨中的碰撞密度，并从理论上计算了碰撞密度的比热关系依赖性，在$50.85 \pm 0.3cm$中用COMSOL Multiphysics计算了热中子在石墨中的扩散长度，在$50.95 \pm 0.5cm$中用MCNP计算了热中子在石墨中的扩散长度。最后用有限元法反迭代计算了中子总截面。
英文摘要：
Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \pm 0.3cm$ using COMSOL Multiphysics and $50.95 \pm 0.5cm$ using MCNP. Finally the total neutron cross-section is derived using FEA in an inverse iteration form.
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