基于Dragon/Donjon的钠冷快堆反应性反馈参数计算与分析

Calculation and Analysis of Reactivity Feedback Coefficients in a Sodium-cooled Fast Reactor using Dragon/Donjon Code

反应性反馈参数是反应堆系统行为与事故安全分析的基础输入参数,与堆芯局部的材料、温度或几何尺寸的变化息息相关,是用于检验堆芯分析程序与方法有效性的关键参数之一。本文针对法国超凤凰钠冷快堆的中子学基准题,基于传统的两步法和Dragon/Donjon程序进行计算,获得一系列的反应性反馈参数,并与利用蒙特卡洛程序Serpent的基准题结果进行对比,分析了中子核反应截面数据库和堆芯计算方法对计算偏差的影响。计算结果表明,Dragon/Donjon程序可用于钠冷快堆反应性反馈参数的计算,各项反应性系数与基准题的偏差可接受。与扩算理论的MCFD算法相比,输运理论的SP3算法对于堆芯径向均匀膨胀、钠密度变化或钠空泡效应、燃料轴向膨胀产生的反应性反馈系数的计算偏差更小。对于不涉及结构材料截面温度效应的反应性系数,基于JEFF3.1.1库和ENDF/B 8.0库的结果差异很小。

Reactivity feedback coefficients are basic input data for systematic behaviors and accidental safety analysis of a nuclear reactor. They’re part of the key results used to verify and validate a reactor physics code and its corresponding methods, because these coefficients are intensely related to the variation of materials, temperature and dimension in the core. A set of whole-core simulations of the Superphenix reactor benchmark was performed with the Dragon/Donjon code using classic two-step scheme, obtaining a series of reactivity feedback coefficients and their differences comparing with the reference calculation using Serpent code. The results show that the Dragon/Donjon code is suitable to calculate the reactivity feedback coefficients of a sodium-cooled fast reactor, and the discrepancies of reactivity feedback coefficients between the Dragon/Donjon code and the benchmark results is acceptable. The results using SP3 method based on transport theory are more accurate than that using MCFD method based on diffusion theory, especially on the reactivity coefficients of reactor radial expansion, sodium density or void effect, axial fuel expansion. For the reactivity coefficients without the Doppler effect of structural materials, there’s not obvious difference observed in the results with JEFF 3.1.1 lib and ENDF/B 8.0 lib.