活化法测量CFBR-II堆中子注量和中子能谱

采用活化法研究了CFBR-II堆中子能谱、中子注量分布和辐照样品对中子场的扰动。建立了用于求解中子能谱的SAND-II解谱程序。对实验结果的分析表明,活化法得到的中子注量率与裂变室得到的结果是一致的,辐照样品对中子能谱有一定的软化。     

Characterization of neutron field in the experimental fast reactor JOYO for fuel and structural material irradiation test

The experimental fast reactor JOYO has been operated as an irradiation test facility for fast reactor fuel and structural material since 1983 with its MK-II core. During this time, an extensive study was conducted to characterize the neutron field in order to assure the accuracy and reliability of neutron fluence. Neutron flux for a given irradiation test was calculated using a core management code system based on three-dimensional diffusion theory. It was then corrected with the adjusted neutron spectrum by means of the multiple foil activation method. The neutron fluence calculation accuracy in the fuel region was evaluated within a 5% error by comparing the burn-up of spent fuel with the measured values, which had been obtained from their post-irradiation examination. At positions away from the fuel region, the neutron flux distribution was calculated using a two-dimensional transport code. A Monte Carlo code was also used to analyze the detailed neutron flux distribution within an irradiation test subassembly that had a heterogeneous internal structure. With the neutron flux results various irradiation parameters, such as displacement per atom (dpa) and helium production, could be evaluated. A helium accumulation fluence monitor has been developed to measure not only neutron fluence but also helium production. Neutron flux and fluence obtained from the core management calculations were compiled as a database for users’ convenience together with related irradiation information and fuel subassembly material compositions. These data are expected to be widely used in the post-irradiation analysis of fuel and structural material.     

中子吸收材料累积中子注量研究

中子注量可作为加速辐照实验的辐照指标。为了通过加速辐照的方式检验中子吸收材料的中子吸收性能,计算了中子吸收材料贮存不同时间下的中子注量。通过对乏燃料组件初始富集度、燃耗深度以及乏池温度、可溶硼浓度的研究,得到中子吸收材料在乏池贮存时中子注量的包络值,同时计算得到不同贮存时间材料¹⁰B的消耗量。结果表明,材料的中子吸收性能在贮存10～60 a的情况下并无明显变化。本文结果可为检验材料的中子吸收性能提供支持。     

Neutron Fluence and Electric Field Strength Dependencies of the Rate of Volume Damage Introduction in Silicon P-N Junctions

The anamolous behavior of neutron-induced defect clusters in the high field space-charge region of silicon p-n junctions is investigated. The space-charge region recombination current variation with neutron fluence and the junction electric field strength present during irradiation is examined and an electric field strength dependence for the defect clusters is found to exist. The space-charge region rate of volume damage introduction is observed to be a function of both neutron fluence and the junction electric field strength present during irradiation. An empirical expresion for the dependence of the rate of volume damage introduction on fluence and junction field strength present during irradiation is developed. Possible mechanisms for the observed field dependence are discussed.     

集成运算放大器的中子辐照损伤效应研究

论文针对双极型运算放大器的中子辐照损伤效应开展实验与理论研究。从双极型器件的中子辐照损伤机理出发,考虑器件电流增益随中子注量的退化,对集成运放的偏置电流、开环增益、共模抑制比与电源抑制比等敏感参数展开试验与理论研究。基于中子辐照损伤系数,针对集成运放中敏感参数随中子注量的退化进行仿真研究,并通过电路敏感性分析明确运放中子辐照损伤的敏感器件与敏感单元。     

能谱变化对活化法中子注量测量的影响

在反应堆中子注量测量中,活化探测器可能会经历多个燃料循环的中子辐照,不同燃料循环的中子能谱也会发生变化。考虑到中子能谱变化的影响,对某批次国产反应堆压力容器辐照材料进行中子注量测量修正。计算结果表明,探测器权重快中子注量率（E＞1.0 MeV）修正后比理论中子注量率（E＞1.0 MeV）高1.75%;与修正前相比降低了3.73%,中子能谱变化的影响不容忽视。      

核石墨耐辐照性能的评价方法

在高温气冷堆运行过程中,作为堆内构件的石墨经受高温和快中子的辐照,会经历先收缩后膨胀的宏观尺寸形变,并在膨胀至原始尺寸时到达使用寿命。在石墨尺寸形变的过程中,石墨内部气孔的结构和数目均有明显变化。当辐照剂量接近使用寿命时,石墨内部气孔数目明显增加,导致其力学性能急剧下降而退出服役。He⁺、C⁺、Xe⁺离子辐照实验表明,在200keV10¹⁴cm^-2Xe⁺离子辐照下,石墨气孔形貌变化明显。这一结果可作为石墨辐照性能的评价方法。     

中子辐照诱导Si PIN光电二极管暗电流增大的数值模拟

分析了中子辐照诱导Si PIN光电二极管暗电流增大现象的机理,建立了Si PIN光电二极管的器件物理模型和中子辐照效应模型。运用MEDICI软件进行数值模拟计算,得出了1MeV中子在辐照注量为1010～1014cm-2时,Si PIN光电二极管暗电流变化的初步规律。数值模拟结果与相关文献给出的实验结果吻合较好。     

300#反应堆单晶硅掺杂质量控制

300^#反应堆在二十世纪80年代开始开展单晶硅掺杂，目前有6根管道用于硅样品辐照。通过对热中子注量计算和辐照中热中子注量监测的控制，得出目标电阻率命中率达到90％以上。采用旋转电机在辐照中旋转硅样品的方法，控制硅样品径向掺杂的不均匀性小于5％，利用管道轴向中子注量率平坦区，并进行适当的位置调整，控制硅样品轴向掺杂的不均匀性小于5％。      

光敏晶体管的中子位移损伤效应研究

通过开展光敏晶体管的反应堆中子辐照实验,获得位移效应实验结果,并分析位移损伤机理。研究发现,在3×10¹¹~5×10¹²cm^-2中子注量范围内,光敏晶体管增益和光响应度的下降导致集电极输出电流下降。增益的倒数与注量的增加呈线性关系,注入电流越大,线性关系的斜率越小。理论分析表明,通过提高基区掺杂水平或减小基区宽度,可提高增益的抗辐射水平;不同反向偏置电压下的初级光电流辐照前基本相同,随着辐照注量的增大,差异逐渐增大,反向偏置电压越大,初级光电流的退化越小;通过采用PIN结构或加大反向偏置电压来展宽耗尽区以减少受位移效应严重影响的扩散电流份额,可提高初级光电流的抗辐射水平。与PIN光电二极管不同,本实验注量范围内,光敏晶体管的暗电流随注量的增大而减小。     

纳米晶Al的快中子辐照效应研究

采用真空热压技术将自悬浮定向流法制得的纳米Al粉压制成平均晶粒尺寸约为120nm的块体,并对其进行了注量为1.9×1012~7.2×1014 cm-2的快中子(E1 MeV)辐照。通过X射线衍射(XRD)分析、扫描电子显微镜与能谱(SEM-EDS)分析和显微硬度测试研究了快中子辐照对纳米晶Al的微观结构和显微硬度的影响。研究结果表明:快中子辐照同时造成了纳米晶Al的平均晶粒尺寸增大和显微硬度提高。随快中子辐照注量的增大,纳米晶Al的平均晶粒尺寸和显微硬度分别增大了2.09%~9.09%和3.54%~4.37%。纳米晶Al的平均晶粒尺寸的增长率随快中子注量的增加而增大。     

微型中子源反应堆中子谱参数测量

以Au、Zr和Fe为活化探测器,采用裸探测器法测量中国原子能科学研究院微型中子源反应堆的中子谱参数f、α、f_F和φ_th。内辐照座的α、f和f_F分别为－0.007±0.003、20.8±0.4、5.5±0.2。该方法对φ_th的测量结果与4πβ-γ符合法的一致,相对偏差小于2%。与SLOWPOKE相比,微堆有较高的α、f_F值。与已有测量数据的比较表明,微堆中子谱在很长一个时期内是稳定的,利用微堆作为中子源的k₀法中子活化分析不需中子注量率监测器,且比较器一经照射和测量后,可用于其后较长时间内所有分析的计算标准。     

反应堆压力容器中子辐照损伤的分析和评价

对于材料已经确定的反应堆压力容器,其辐照脆化效应的主要因素是快中子积分通量。本文应用中子输运格林函数法验算了秦山核电站压力容器1/4厚度处最大快中子通量。分析和评价结果表明,该压力容器的设计对中子辐照是安全的。     

Low temperature fast-neutron and gamma irradiation of glass fiber-epoxy composite. Part 2: Structure and chemistry

Structure and chemistry of S2-glass fiber-epoxy composite to fast-neutron and gamma irradiation at 4.2 K have been studied. Fast-neutron fluence ranged from 0.9 × 10²² to 1 × 10²³ n/m². At the neutron fluence of 3.1 × 10²² n/m², ordering of the amorphous S2-glass structure intervened. Ordering was followed by partial crystallization at the neutron fluence of 1 × 10²³ n/m². Although the epoxy matrix retained its amorphous structure, reactor irradiation induced crosslink formation and hydrogen evolution. These processes became significant at neutron fluences equal to or higher than 1.8 × 10²² n/m². No evidence has been found for hydrogen bubble or void nucleation at the glass-epoxy interface at the neutron fluence of 3.1 × 10²² n/m². Futhermore, sharp compositional changes measured at the glass-epoxy interface preclude an extended irradiation-induced atom mixing at this neutron fluence.     

Mechanical modeling of creep, swelling and damage under irradiation for polycrystalline metals

A constitutive equation of creep, swelling and damage under irradiation for polycrystalline metals applicable to structural analyses in multiaxial state of stress is developed. After reviewing microscopic mechanisms of irradiation creep and swelling, the relevant theories proposed so far from the view point of metallurgical physics and their applicability are discussed first. Then a constitutive model is developed by assuming that creep under irradiation can be decomposed into irradiation-affected thermal creep and irradiation-induced creep. By taking account of the Stress-Induced Preferential Absorption (SIPA) mechanism, the irradiation-induced creep is represented by an isotropic tensor function of order one and zero with respect to stress, which is, at the same time, the function of neutron flux and neutron fluence. The volumetric part of the irradiation-induced creep is identified with swelling. The irradiation-affected thermal creep is described by modifying Kachanov-Rabotnov theory for stress-controlled creep and creep damage by incorporating the effect of irradiation. Finally irradiation creep and swelling of 20% cold-worked type 316 stainless steel at elevated temperature are predicted by the proposed constitutive equations, and the numerical results are compared with the corresponding experimental results.     

离体人血中子吸收剂量的估算

本文介绍了如何设计小体积离体人血样品的中子照射实验,以及计算中子注量、吸收剂量的方法,并给出了几个参考数据。     

中子辐照导致线阵电荷耦合器件电荷转移效率退化实验研究

利用TRIGA型脉冲反应堆提供的快中子，对线阵电荷耦合器件进行中子辐照实验研究。研究结果表明：在10¹²～10¹³cm^-2中子注量范围内，该器件的电荷转移效率（CTE）随辐照中子注量的增加而线性下降；电荷转移效率的下降与电荷包在沟道中的转移时间及转移电荷包的电量有关。      

Calculation of neutron fluence in the region of the pressure vessel for the history of different reactors by using the Monte-Carlo-method

Embrittlement of pressure vessel material caused by neutron irradiation is a very important problem for VVER-440 reactors. For the estimation of the fracture risk highly reliable neutron fluence values are necessary. For this reason a special theoretical determination of space dependent neutron fluences has been performed mainly on the basis of Monte-Carlo calculations. The described method allows the accurate calculation of neutron fluences near the pressure vessel in the height of the core region for all reactor histories and loading cycles in an efficient manner. To illustrate the accuracy of the suggested method a comparison with experimental results was done. The calculated neutron fluence values can be used for planning the loading schemes of each reactor according to the safety requirements against brittle fracture.     

活化法测量散裂靶中子能谱的实验验证

散裂靶中子的能谱对加速器驱动次临界系统的倍增因数和嬗变率等影响很大,计算表明散裂靶中子谱在MeV能区与裂变中子谱相近。本文利用活化法测量临界装置的泄漏中子谱和中子注量率,提出了用In、Al、Mg、Ti、Au、Zn、Ni、Rh、Fe和Co等活化箔测量散裂靶中子能谱和中子注量率的方案。结果表明,将活化箔在散裂靶中子场中辐照5h,中子注量最高达5×1014 cm-2量级,辐照后1h内取出活化箔,根据半衰期的长短安排测量顺序,可测量散裂靶的中子能谱和中子注量率。