Evaluation of DNBR with neutronics calculation in LWR systems

S. Khan, Umasankari Kannan
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Abstract

The heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.
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在低水堆系统中进行中子计算的DNBR的Evaluation
从排热角度出发,利用轻水堆系统的热流密度来估计系统的离核沸腾比(DNBR),这是核反应堆重要的热工水力参数。DNBR表示一个操作安全限制即核电站必须从指定的操作有足够的保证金DNBR限制保证它的安全。DNBR的评估使用热工分析代码,并使用来自中子计算的输入。本文提出了标准中子计算中最小DNBR (MDNBR)的计算方法。利用核心物理分析程序进行了DNBR的计算,并研究了全循环长度下MDNBR的燃耗变化。计算的结果提出了使用2700 MWth / 900兆瓦的平衡核心印度压水反应堆(IPWR)。计算使用VISWAM-TRIHEXFA代码系统执行。利用点阵分析程序VISWAM生成了IPWR的小群点阵参数库。利用岩心分析代码TRIHEXFA对平衡岩心结构进行了岩心跟踪计算。在三维有限差分岩心仿真工具TRIHEXFA中引入了一阶热液反馈模型。利用TRIHEXFA中实现的W-3 Tong和OKB-Gidropress相关性进行了DNBR估计所需的临界热通量计算。
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