Measurement and analysis of fission rates in a spherical mockup of uranium and polyethylene

Measurements of the reaction rate distribution were carried out using two kinds of Plate Micro Fission Chamber （PMFC）. The first is a depleted uranium chamber and the second an enriched uranium chamber. The material in the depleted uranium chamber is strictly the same as the material in the uranium assembly. With the equation solution to conduct the isotope contribution correction, the fission rate of 238U and 235U were obtained from the fission rate of depleted uranium and enriched uranium. Then, the fission count of 238U and 235U in an individual uranium shell was obtained. In this work, MCNP5 and continuous energy cross sections ENDF/BV.0 were used for the analysis of fission rate distribution and fission count. The calculated results were compared with the experimental ones. The calculation of fission rate of DU and EU were found to agree with the measured ones within 10% except at the positions in polyethylene region and the two positions near the outer surface. Because the fission chamber was not considered in the calculation of the fission counts of 238U and 235U, the calculated results did not agree well with the experimental ones.     

Measurement and analysis of the 238U(n, 2n) reactionrate in depleted uranium/polyethylene shells

In order to check the conceptual design of the subcritical blanket in a fusion-fission hybrid reactor, a depleted uranium/polyethylene simulation device with alternate shells has been established. The measurement of the ²³⁸U(n, 2n) reaction rate was carried out using an activation technique, by measuring the 208 keVγ rays emitted from ²³⁷U. The self-absorption of depleted uranium foils with different thicknesses was experimentally corrected. The distribution of the ²³⁸U(n, 2n) reaction rate at 90° to the incident D⁺ beam was obtained, with uncertainty between 5.3% and 6.0%. The experiment was analyzed using MCNP5 code with the ENDF/BVI library, and the calculated results are all about 5% higher than the measured results.     

D-T中子诱发贫化铀球壳内裂变率分布实验

中子诱发裂变反应率是表征和检验中子在材料中的输运、裂变放能等过程的重要物理量.贫化铀球壳裂变反应率径向分布数据,可为铀核数据宏观检验及研究裂变放能与贫化铀球壳厚度的关系提供数据支持.本文设计了内径为13.1 cm,外径分别为18.10,19.40,23.35,25.40,28.45 cm的五种不同厚度的贫化铀球壳组合装置;利用位于球壳中心的氘氚中子源轰击贫化铀球壳装置,中子产额约为3×10~(10)—4×10~(10)s~(-1);在"赤道"平面与入射氘束成45°方向测量裂变反应率随径向分布的情况.为了克服裂变室和俘获探测器等自身对模型和中子场的扰动,本文选择与装置材料相同的贫化铀材料作为活化探测器,以活化探测器中的裂变碎片143Ce发射的γ射线作为测量对象,通过HPGe探测器测量的γ射线数,基于~(143)Ce裂变产额数据反推裂变反应率.通过实验获得了贫化铀球壳内的裂变率及其径向分布规律,裂变反应率和相对标准不确定度分别位于5.28×10~(-29)—7.58×10~(-28)之间和6%—11%之间.基于蒙特卡罗程序和ENDF/BVI.8数据库完成了模拟计算,并与实验结果进行了对比分析,两者在不确定度范围内一致.     

热处理对氢气放电实验用靶的影响

对本底靶进行热处理后用TiT源重新开展打靶实验,未知谱线消失,证明未知谱线不是由靶中杂质造成的。将同种材料制成的两块本底靶分别进行热处理和不做任何处理,置于氢气放电源中进行照射生成源照靶,用TiT源打这两块源照靶发现照射前经热处理的靶产生的4条未知谱线较强,证明热处理能够增强靶吸附未知粒子的能力。     

基于白光中子源的197Au中子辐射俘获截面测量及共振参数分析

中子辐射俘获反应在反应堆运行、核装置设计及核天体物理研究中起重要的作用. 4πBaF₂探测装置有着高时间分辨能力、低中子灵敏度、高探测效率等优点,适合开展中子辐射俘获反应截面数据的测量.中国原子能科学研究院核数据重点实验室建立了伽马全吸收装置(Gamma total absorption facility, GTAF),该装置用28块六棱BaF₂晶体和12块五棱BaF₂晶体构成了外径25 cm,内径10 cm的球壳,覆盖了95.2%的立体角.利用GTAF在中国散裂中子源Back-n束线上,测量了197Au(n, γ)的反应截面数据.测量数据通过能量筛选、PSD方法、晶体多重性筛选进行了初步本底扣除,随后结合对^natC及空样品的测量数据对本底进行了分析及扣除,获得了197Au俘获反应的产额,利用SAMMY程序拟合得到了¹⁹⁷Au在1—100 e V的共振能量、中子共振宽度和伽马共振宽度参数.实验测量结果与ENDF/B-VIII.0数据库符合良好,其共振参数存在一定差异,分析原因可...     

打靶谱测量中的表面效应

通过电子、X射线混合源和X射线源打源照镁靶谱的对比实验,论证了放射源打靶谱的测量中存在着表面效应。它除和电子打靶有关外,也反映了靶表面的一些特性。为此,做了放电源打源照镁靶谱随存放时间变化的实验。实验结果表明:源照镁靶表面可能存在一层阻挡层,阻止新态氢原子向外扩散。根据打磨和未打磨源照镁靶打靶谱的对比实验,推测靶表面的阻挡层可能是一层新态氢原子富集层。为此,又做了镁铼靶打靶谱的对比实验,照射打磨和未打磨镁靶打靶谱的对比实验,根据实验结果,估计新态氢原子富集层可能是一层以氧为核的氢原子和新态氢原子团。由此推测阻挡层是一层以氧为核的新态氢原子团组成的新态氢原子富集层。这种新态氢原子富集层产生了打靶谱测量中的表面效应。     

中国散裂中子源反角白光中子束流参数的初步测量

中国散裂中子源(CSNS)已于2018年5月建设完工,随后进行了试运行.其中的反角白光中子束线(Back-n)可用于中子核数据测量、中子物理研究和核技术应用等多方面的实验.本文报道对该中子束的品质参数测量实验过程以及最终实验结果.实验主要采用中子飞行时间法,利用~(235)U,~(238)U裂变室和~6Li-Si探测器测量了中子能谱和中子注量率,又利用闪烁体-互补金属氧化物半导体探测系统测量了中子束斑的剖面,得到了该束线的初步实验测量结果.其中白光中子的全能谱测量范围eV—100 MeV,给出了不确定度分析;给出了中子注量率两个实验厅位置的满功率值;给出了白光中子在直径60 mm情况下的全能区束斑.通过与模拟结果的比较探讨了以上结果的合理性,并提出了改进计划.这些实验结果为以后该束线的核数据测量和探测器标定实验奠定了基础.     

Measurement and analysis of the neutron-induced total cross-sections of 209Bi from 0.3 eV to 20 MeV on the Back-n at CSNS

The neutron-induced total cross-section of ²⁰⁹Bi is crucial for the physical design and safety assessment of lead-based fast reactors, and the quality of experimental data should be improved for evaluation and application.A recent experiment was conducted on the back-streaming white neutron beamline(Back-n) at the China Spallation Neutron Source(CSNS) using the neutron total cross-section spectrometer(NTOX). The neutron energy was determined using a fast multi-cell fission chamber and...