Study of safety performance of the 241Am fire alarm source

The safety performance of ^241Am fire alarm sources mde by using powder metallurgical technology has been preliminarily studied ,so as to determine and allowable maximum energy limit value of the alpha particles outgoing from this kind of sources in light of radiation safety and the present technology,The yielded results show that ^241 Am leak has not been found when the peak energy of the alpha energy spectrum of this kind of soureces is less than 4.96MeV.     

RELATIVE CALIBRATION OF ~(241)Am FIRE ALARM SOURCE

According to the requirements of improving production of Am source used for smoke-sensitice fire alarm making, a method and equipment of its fast relative calibration have been developed. Six measuring equipments have being successfully applied at production lines of 241Am fire alarm source and factories of automatic smoke-sensitive fire alarm making.     

Influence of some factors on alpha energy spectrumof 241Am fire alarm source

Several primary factorsinfluencing the alpha energy spectrum of ²⁴¹Am fire alarm source have beenstudied in order to get betteralpha energy spectrum. The results show that the homogeneityandthe thickness of metal surface coat and the size of active area of thesource haveconsiderable influence on the alpha energy spectrum of thesource.     

数字高斯脉冲成形算法仿真研究

高斯脉冲具有良好的时间、频率响应和较高的信噪比,探测器输出信号通常被滤波成形为高斯波形或类高斯波形。在高斯脉冲成形算法基础上,引入截止频率和品质因子,讨论成形参数对成形脉冲幅度、宽度的影响。通过与梯形脉冲成形算法对比,研究高斯脉冲成形算法的滤波效果以及对核信号幅度谱的作用。结果表明,在相同达峰时间条件下高斯脉冲成形算法具有更好的噪声抑制能力,同时可减小核信号幅度谱的半高宽;结合使用梯形脉冲成形算法可改善高斯脉冲成形算法在处理堆积脉冲中的不足,提高核信号通过率。     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.