不同评价核数据库压水堆包壳材料对反应性的影响及分析

随着各国新版本评价核数据库的发布,不同版本评价核数据库对反应性的影响并不完全一致,为选取高精度的评价核数据库,以几何尺寸、核素种类较简单的压水堆包壳材料为研究对象,基于新版本的评价核数据库ENDF/B-VII.0、JEFF-3.3、JENDL-4.0和CENDL-3.1,采用NJOY2016程序制作压水堆常见包壳材料(不锈钢包壳、铝包壳和锆包壳)的截面数据。通过组件程序DRAGON5.0.1挂载不同评价核数据库版本得到包壳材料的多群截面库,计算WIMS库更新计划(WLUP)系列临界基准题,并将计算结果与实验值进行比较。结果表明,52Cr、56Fe、90Zr、91Zr、92Zr和94Zr这6个核素在不同评价核数据库版本中对反应性影响均较大;采用CENDL-3.1和JENDL-4.0这2个版本评价核数据库制作的压水堆包壳材料,其计算结果与实验值较为接近。

基于等价理论的压水堆包壳材料共振计算方法研究

共振自屏效应的处理是影响压水堆组件程序反应性精度的主要因素之一,压水堆锆包壳材料同样具有共振自屏效应,忽略其影响会对反应性造成100~300 pcm(1 pcm=10-5)的偏差。目前,主要通过提供经验上的参考稀释截面与包壳等价理论处理包壳材料的共振自屏效应,但并未对其适用性及精度进行完整的分析。因此,本文采用DRAGON程序,通过一系列压水堆算例对这2种方法进行测试,确定包壳共振自屏效应的主要影响因素以及这2种方法的适用性。结果表明,包壳材料的共振自屏效应仅仅与包壳区的原子核密度、厚度、慢化区的水铀比有关,并且参考稀释截面方法可以满足大部分典型压水堆系统的计算精度,但是对于包壳区尺寸、原子核密度、慢化区水铀比变化较大的系统计算精度较差,而包壳等价理论计算精度和普适性强,可用于不同类型压水堆系统包壳材料的共振自屏计算。

Revisiting Stainless Steel as PWR Fuel Rod Cladding after Fukushima Daiichi Accident

In the past, stainless steel was utilized as cladding in many PWRs （pressurized water reactors）, and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as cladding material mainly due to its lower neutron absorption cross section. Now, stainless steel cladding appears as a possible solution for safety problems related to hydrogen production and explosion as occurred in Fukushima Daiichi accident. The aim of this paper is to discuss the steady-state irradiation performance using stainless steel as cladding. The results show that stainless steel rods display higher fuel temperatures and wider pellet-cladding gaps than Zircaloy rods and no gap closure. The thermal performance of the two rods is very similar and the neutron absorption penalty due to stainless steel use could be compensating by combining small increase in U-235 enrichment and pitch size changes.