芪探测器响应效率各向异性的计算

在Linux操作系统上使用PEGS4程序离线计算了光子及电子与芪晶体和铝的相互作用的截面数据库,并利用计算的截面数据库作为该系统中的Dosnrc程序的输入数据计算得到了芪晶体对光子各能量点的响应函数.并对来自与探测器轴线成不同角度不同能量的平行光束的探测效率进行了计算,通过测量标准γ源22Na的反冲电子谱和探测效率来与计算的效率和响应函数进行比较.     

快堆用蒙特卡罗多群数据库制作方法研究

基于离散角方法,开发了蒙特卡罗多群数据库生成程序MGXSMC,该程序可以实现从输入文件读取截面数据或者从指定格式的截面库中读取截面,产生可供蒙特卡罗程序MCNP或RMC计算的数据库,并且可自动生成相应的索引文件列表。采用二维两群不带反射层的国际原子能机构(IAEA)压水堆(PWR)基准题和铅基快堆(RBEC-M)基准题对MGXSMC程序加工产生的核数据进行验证,计算结果表明,采用P₅阶近似多群截面与连续点截面计算的有效增殖系数(k_eff)结果相差24 pcm(1pcm=10^-5),而采用P₀阶近似多群截面与连续点截面计算的k_eff结果相差较大。由此说明蒙特卡罗多群数据库的制作方法和所开发的程序是正确的,同时,中子各向异性散射对铅基快堆计算结果影响较大,故制作蒙特卡罗多群数据库时应加入中子散射角数据。     

二维输运燃耗耦合程序开发与验证

为准确计算反应堆内燃耗问题,建立了基于二维离散纵标法及BATEMAN燃耗方法的输运燃耗耦合计算方法,并开发相应的计算程序。基于ENDF/B-Ⅶ评价库开发了175群中子和42群光子截面数据库MUSE-F1.0,采用OECD/NEA发布的MOX燃料快堆基准题对耦合计算方法及程序系统进行验证计算。结果表明,耦合计算程序结果与基准题吻合良好,误差在8%以内,初步验证了耦合计算程序在快堆嬗变工程应用中的可行性。     

NaI（Tl）晶体对γ射线响应函数的蒙特卡罗模拟

以137Cs放出能量为0．662MeV的7光子在NaI（T1）晶体中的响应函数为研究对象,探讨响应函数的计算方法。对能量沉积谱分段,计算出落在每一能量段内的概率,依此将实验测得的参数代入程序中对谱线进行拓展形成符合高斯分布的全能峰。用蒙特卡罗方法建立物理模型,利用设计的模拟软件模拟出整个光子的运动过程,同时显示出晶体探测到的γ光子响应函数。     

基于RELAP5和FLUENT的多尺度耦合分析方法研究

基于二次开发得到的铅冷快堆一维系统程序RELAP5_LEAD和三维计算流体力学程序FLUENT,利用动态链接库技术和FLUENT用户自定义函数,开发了多尺度耦合分析程序RELAP5/FLUENT。在单相范围内,分别利用耦合程序RELAP5/FLUENT开展简单铅冷串联管道的瞬态流动和传热模拟、简单铅冷闭式回路的瞬态流动模拟,并与RELAP5_LEAD计算结果开展Code-to-Code对比分析。研究结果表明,RELAP5/FLUENT计算结果与RELAP5_LEAD模拟结果吻合良好,耦合程序的开发取得了初步成功,可用于分析铅冷快堆堆内的复杂三维热工水力现象。     

NaI(T1)晶体对γ射线响应函数的蒙特卡罗模拟

以137Cs放出能量为0.662MeV的γ光子在NaI(T1)晶体中的响应函数为研究对象,探讨响应函数的计算方法.对能量沉积谱分段,计算出落在每一能量段内的概率,依此将实验测得的参数代人程序中对谱线进行拓展形成符合高斯分布的全能峰.用蒙特卡罗方法建立物理模型,利用设计的模拟软件模拟出整个光子的运动过程,同时显示出晶体探测到的γ光子响应函数.     

Q值在线计算程序QCALC

Ｑ值计算程序ＱＣＡＬＣ及其帮助文件已被开发移植并建立在ＭＩＣＲＯ－ＶＡＳ－Ⅱ计算机上。该程序系统可以在线计算各种核反应的Ｑ值和阈能以及各种核衰变的Ｑ值，并可给出这些计算结果的误差。该程序还能在线检索１９９３年评价的原子质量数据库。     

蒙特卡罗计算程序高温点截面数据的增补

MCNP程序是目前在反应堆计算分析中广泛应用的蒙特卡罗计算程序。MCNP自带的截面库中的绝大多数核素只包含21℃（294K）下的截面数据。虽然也提供了一些常用核素的高温点截面数据，但温度点划分很粗，间隔跨度很大，如300、400、500、587和600K等温度下的核截面数据。     

SRAC程序在中国先进研究堆上的应用研究

SRAC程序是由日本原子能研究机构发布的在Linux系统下运行的堆芯物理计算程序包,包括栅元计算程序、堆芯计算程序和燃耗计算程序.本程序采用基于JENDL或ENDF/B系列的数据库,可处理300多种核素的截面参数.该程序具有计算速度快、堆芯燃料管理方便等优点.本文对SRAC程序在中国先进研究堆(CARR)上的应用进行了初步研究,利用SRAC程序对CARR进行了临界计算和首炉燃耗计算.通过与WIMS-CITATION程序系统计算结果的比较,初步确定了SRAC程序是CARR在线计算工具很好的选择,并为今后的实际应用奠定了良好的技术基础.     

核数据处理软件NECP-Atlas中的光子相关数据计算方法研究

NECP-Atlas是西安交通大学自主研制的核数据处理软件,具有丰富的功能,可将评价核数据制作为后续核设计所需的应用核数据库,本文在NECP Atlas中建立了光子相关数据的计算方法,可计算产生中子核反应释放的瞬发光子产生截面、光子与原子的反应截面,裂变产物衰变释放的缓发光子多群产生矩阵,以及光子辐照损伤截面等数据。数值结果显示,如果不考虑缓发光子,钠冷快堆中控制组件、反射层组件的光子功率与参考解的最大偏差可达3258%、2041%,采用NECP Atlas计算的多群缓发光子产生矩阵后两类组件偏差降为093%以下。采用文献结果对Fe的光子辐照损伤截面进行了验证,计算结果与参考解吻合良好。     

Validation and upgrading of the recommended cross-section data of charged particle reactions: Gamma emitter radioisotopes

An upgrade and validation test of the recommended cross-section database for production of gamma emitter radioisotopes by charged particle induced reactions, published by the IAEA in 2001, was performed. Experimental microscopic cross-section data published earlier or measured recently and not yet included in the evaluation work were collected and added to the primary database in order to improve the quality of the recommended data. The newly compiled experimental data in general supported the previous recommended data, but in a few cases they influenced the decision and resulted in different selected cross-section data sets. A Spline fitting method was used to calculate the recommended data from the selected data sets. Integral thick target yields were deduced from the newly calculated recommended cross-sections and were critically compared with the available experimental yield data.     

D-T中子照射下贫化铀球、钒球介质内中子能谱和伴生γ能谱测量

为获得介质内中子能谱及伴生γ能谱的实验数据，在中心D-T中子照射下，用18mm×20mm的茋闪烁体探测器，测量了与D+束成45°角的水平方向距球心7、10、13、16、19、22cm位置处贫化铀球介质内的中子能谱和伴生γ能谱，以及钒球内与D+束成0°角、距离球心1.8、4.8和8.3cm处的中子能谱和伴生γ能谱。用MCNP/4B程序和ENDF/B-VI库数据对实验模型进行模拟计算，并与实验结果进行了比较。     

Development of large-area composite stilbene scintillator for fast neutron detection

Fast neutron applications have gained popularity with the growth of fast neutron production facilities. Covering a larger area and/or wider angle can be one of the advantages of a fast neutron detector. In the present study, a large-area composite stilbene scintillator with the dimensions of 200 mm (D) × 20 mm (H) was fabricated to examine its scintillation properties and to evaluate its applicability to fast neutron detection. The detector response of small- and large-area composite stilbene scintillators for neutrons and gamma rays was measured and compared with that of commercial and small single-crystal stilbene scintillators. To this end, the response of each scintillator was measured for radioisotopes as well as mono-energetic neutrons generated by a Tandem accelerator. The neutron–gamma separation performance of the large-area composite stilbene scintillator was evaluated in terms of figure-of-merit (FoM) using the digital pulse shape discrimination method. The composite stilbene scintillator showed good energy linearity, as determined from its recoil proton spectra, with reasonable n–γ separation capability. The results indicated that the composite stilbene scintillator could be applied to the field of fast neutron detection, especially when a large area and/or a wide angle is to be covered and could be a good alternative to liquid scintillators.     

核截面库可视化研究

对ENDF/B-6.8数据库的点截面形式(ACE格式)数据进行了详细的解析,以此为基础编写程序实现了对于ACE格式数据的读取与相关处理。本程序实现了对于ENDF/B-6.8数据库核截面-能量的绘图、缩放、比较等可视化功能,还提供了感兴趣核数据的标准图片格式和/或标准文本格式输出。通过程序输出数据与参考数据的比较,对程序结果的正确性进行了适当的验证。     

An evaluation of neutron and gamma-ray-production cross-section data for lead

A survey was made of the available information on neutron and gamma-ray-production cross-section measurements of lead. From these and from relevant nuclear-structure information on the Pb isotopes, we prepared recommended neutron cross-section data sets for lead covering the neutron energy range from 0.00001 eV to 20.0 MeV. The cross sections are derived from experimental results available to February 1972 and from calculations based on optical-model, DWBA, and Hauser-Feshbach theories. Comparisons which show good agreement between theoretical and experimental values are displayed in a number of graphs. Also presented graphically are smoothed total cross sections, Legendre coefficients for angular distributions, and a representative energy distribution of gamma rays from resonance capture.     

GdI3:Ce闪烁体中子探测性能测试与改进研究

为改善GdI₃:Ce闪烁体在探测中子过程中的γ抑制能力,使用Geant4和XCOM计算了其γ线性吸收系数,并通过模拟计算与实验测量研究了铅屏蔽法抑制γ的有效性。结果表明：GdI₃:Ce闪烁体在探测中子过程中易受低能γ射线的干扰;随着铅层厚度的增加,100 keV~1 MeV的γ射线对中子探测的干扰减小,而3~10 MeV的γ射线的干扰呈先增加后减小的趋势。对²⁵²Cf中子源的实验测试发现,在碘化钆闪烁体外围添加铅层后,中子峰得以显现;随着铅层厚度的增加,中子峰内净计数减小,而净计数与本底计数的比值上升。模拟和实验结果均表明,在使用GdI3:Ce闪烁体探测中子时,应根据中子探测效率和信噪比的优化确定γ屏蔽铅层的厚度。     

基于3层硅探测器的主动式中子个人剂量计研制

研制了一种具有能量分辨能力的中子个人剂量计。该个人剂量计由3层硅探测器组成,硅探测器周围装有⁶LiF、聚乙烯和含硼聚乙烯作为转化体、慢化体和吸收体。个人剂量计有直读和解谱两种工作模式。直读模式将中子能区划分为低能、中能和快中子3个能区,可实时测量。解谱模式可获得快中子能区的中子能量分布。利用GEANT4程序模拟了7 MeV γ射线的能量沉积谱,设置适当的甄别域以降低γ射线的影响。采用GEANT4程序模拟了个人剂量计对不同能量入射中子的个人剂量当量率响应。在加速器单能中子参考辐射场中完成了单能中子剂量响应的实验校准,对模拟计算的响应函数进行了实验修正,并得出了不同能区的平均中子个人剂量当量率响应。     

圆柱氧化钍装置上的中子及伽马穿透能谱实验测量

利用D-T中子对氧化钍圆柱装置进行了辐照。样品直径为30 cm、厚度为15 cm。在不同的布局状态下,采用芪晶体闪烁体基于反冲质子法对1 MeV~15 MeV中子能谱进行了实验测量;基于反冲电子法对中子辐照下和辐照后的0.5 MeV以上伽马能谱进行了测量。对能谱中的分布规律、与布局状态的关系进行了比对分析。     

Calculation of the cross-sections for fast neutrons and gamma-rays in concrete shields

This work is concerned with the theoretical calculation of both the total mass attenuation coefficients (μ/ρ) for gamma rays and the effective removal cross-sections (Σ_R) for fast neutrons. Calculations were carried out for four types of concrete of different densities namely (1) dolomite–sand ρ=2.5 g.cm⁻³, (2) barite–barite ρ=3.49 g.cm⁻³, (3) magnetite–limonite ρ=3.6 g.cm⁻³ and (4) ilmenite–ilmenite ρ=3.69 g cm⁻³. The total mass attenuation coefficients have been calculated at energies from 10 keV to 1 GeV using the XCOM (version 3.1) computer program and cross- section database for elements from Z=1 to 100. Also the effective removal cross-section have been calculated using the elemental composition of the concrete mixes. The calculated values are widely needed and used as a database for radiation shielding design of research reactors, power station and particle accelerators. The results were displayed in tables and figures. Calculated results were compared with those previously measured and a reasonable agreement was found between them.     

On the use of silicon as thermal neutron filter

A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 μeV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given.