基于VHTRC的棱柱式高温气冷堆核设计程序验证北大核心CSCD

高温气冷堆是国际公认的固有安全性高的反应堆堆型。针对高温气冷堆包覆颗粒燃料引入的燃料组件的双重非均匀性以及棱柱式堆芯布置的非均匀性和强空间耦合效应,提出基于蒙特卡罗均匀化-确定论输运方法的RMC-SaraGR程序系统作为棱柱式高温气冷堆的核设计程序。基于日本棱柱式高温气冷堆临界实验装置VHTRC基准题,针对此套核设计程序系统开展了均匀化模型研究和初步验证。研究结果表明,基于蒙特卡罗均匀化方法,采用全堆模型、合适的能群结构和分区方式产生组件群常数,并经过超级等效均匀化方法进行等效均匀化修正,可以保证堆芯多群均匀计算具有较高的计算精度。

Validation of a nuclear code system for prismatic high temperature gas-cooled reactors based on the Very High Temperature Reactor Critical Assembly benchmark

The inherent safety characteristic of high temperature gas-cooled reactor (HTGR) is universally acknowledged. On account of the double-heterogeneity of the coated fuel and the irregular core pattern of the prismatic HTGR, a code system based on Monte-Carlo homogenization and deterministic transport concept was used as a nuclear design code system for prismatic HTGR. An investigation on homogenization model and preliminary validation of this code system is conducted based on VHTRC benchmark which is a prismatic HTGR criticality assembly. The validation results indicate that group constants generation based on Monte-Carlo homogenization and appropriate energy group, then modified by super homogenization (SPH) method further, ensures high precision of homogeneous core calculation.