医院中子照射器I型堆热中子束流孔道等效平面源的模拟计算

采用蒙特卡罗程序MCNP模拟计算了医院中子照射器Ⅰ型堆(IHNI-1)热中子束流孔道出口处的等效平面源.对B堆芯进行了临界搜索计算,模拟计算了热中子束流孔道及出口处中子、γ的束流参数,应用等效平面源模型建立了BNCT等效中子、γ平面源.为人体头颅等效模型剂量分布的快速计算提供了较为可靠的平面源.     

轴心受压钢柱抗火设计的优化分析

基于"双重优化设计"概念给出了轴心受压钢柱抗火优化设计方案,利用有限元软件ANSYS的APDL语言及其优化设计模块对其进行了截面尺寸和抗火保护优化设计。分析中采用了两种优化设计方案,即在常温和一定临界温度条件下对轴心受压钢柱进行截面尺寸及抗火保护优化设计。通过对两种设计方案的优化结果进行比较和分析,得到了高温下轴心受压钢柱的抗火优化设计结果随温度变化的规律,给出了考虑经济成本条件下如何选择更加经济合理的优化设计方案的建议。分析结果表明,给出的优化方案可以得到安全、经济、可靠的设计结果,能够节省钢材用量和防火保护材料用量。     

大型燃煤发电机组节油点火方案的设计与比较

锅炉采用节油点火方案是节约燃油的有效手段之一。以某发电厂2×660MW超超临界燃煤机组为例,在对不同节油点火方案进行详细论述的基础上,对锅炉点火系统进行了设计优化,分析对比了"单层微油点火方案"、"单层等离子点火方案"和"两层等离子点火方案"的技术特点。从经济性、适用性和可靠性等方面选出适合本工程特点的点火系统设计方案:推荐设置两层等离子燃烧器(一运一备),取消锅炉燃油系统。     

FAST望远镜外环曲梁的结构方案优化分析

FAST望远镜的AB转轴机构相当于具有两个水平正交轴的万向节,用于调整馈源舱内设备的指向角。其外环(A环)曲梁跨度大,承载重,且要求轻质和高刚度。基于弯扭耦合的曲梁理论建立了A环结构的基本微分方程,给出了结构变形挠度与结构参数之间的关系。以结构总重量作为结构优化总目标,以不同倾角下曲梁跨中最大扰度及最大扰度差作为主要约束条件,应用信赖域法对实腹式箱梁和桁架式梁两种结构方案进行了优化计算,并对两种结构方案进行了比较。初步的优化结果表明桁架式梁结构方案具有更轻的结构重量。     

Al掺杂纤锌矿ZnO的电子结构研究

采用基于密度泛函理论(DFT)的总体能量平面波超软赝势法,电子交换关联能结合广义梯度近似(GGA)PW91形式,对Al原子不同掺杂位置的ZnO超原胞进行了几何优化,计算分析了晶体结构参数及掺杂模型的电子结构.结果表明,几何优化后掺杂晶胞z轴方向出现收缩现象,两种模型在杂质掺入后半导体ZnO导电能力提高,形成置换杂质模型的可能性要大于间隙杂质模型,掺杂Al和基体ZnO的原子数配比直接影响导电性能,最后给出了电子轨道跃迁规律.     

复杂价值工程(CVE)在高速公路路基施工中的应用

本文基于复杂价值工程(CVE)的原理和方法,来开展高速公路建设设计施工优化分析,并基于一个案例:高速公路基础厚度大于等于1米的两套施工方案进行比较研究,最后运用价值工程理论对两套方案的价值系数进行比较研究,得出其中一个较好的优化方案,从而达到降低工程成本及保证施工质量和进度的目的。     

基于ANSYS的挂壁机面板与面板体装配卡扣装配过程分析

针对挂壁机面板与面板体装配时存在卡扣紧的问题,对卡扣装配过程进行了仿真受力分析,提出了两种优化方案;使用ANSYS有限元仿真软件进行相关模拟验证,并对最终优化方案进行评估分析。研究表明,该优化方案可提高装配效率,对结构设计工作具有一定的指导作用。     

基于双目标传感器分布优化的转向架构架状态监测

研究提出一种基于传感器分布优化的转向架构架状态监测方法,采用小波变换和主成分分析对构架正常状态下的多工况振动信号进行特征提取和降维,以支持向量数据描述(SVDD)超球体半径定义双目标优化函数:传感器数量和超球体聚类指标,并基于改进的非劣分层多目标遗传算法(NSGA-II)对传感器分布进行优化。在此基础上,搭建转向架构架状态监测实验平台进行构架异常状态识别研究,结果表明:1)经优化后的传感器分布方案能以较少的数量保证很好的构架异常状态监测效果,当传感器分布优化方案中传感器数量分别为1,2和3时,识别率分别为79.24%,94.80%和99.81%;2)以转向架构架正常状态样本集构建的SVDD模型对异常状态具有很好的识别效果。     

引力波探测望远镜超低热变形桁架支撑结构设计技术

本文针对引力波探测望远镜超低热变形要求,提出了通过设计CFRP铺层改变材料热胀系数的方法,进而解决桁架支撑结构热变形问题。同时针对望远镜装调性能的要求,给出了结构的分段式设计方案。文章首先分析了CFRP的优势、现有的消热化的方式和国内外研究进展,确定了以CFRP为关键材料的三杆式望远镜的方案,并提出了设计指标。其次建立了“材料-热变形、桁架结构-热变形”数学模型,针对材料铺层和结构设计进行优化,给出了优化方案。然后将CFRP材料应用于支撑结构中,采用分段式主支撑结构设计方案,降低了结构加工装调的难度,最后对整体结构进行分析。分析结果表明：机械性能方面,结构基频和最大重力卸载变形满足主支撑结构的要求;热变形方面,基于CFRP铺层特定优化设计的结构方案热变形为普通铺层方案的27.15%、Invar材料支杆方案的6.42%、SiC支杆方案的11.50%和钛合金支杆方案的3.21%,即优化设计可显著降低结构热变形。     

基于Ansys的纸浆模塑模具加热板温度场分析及优化

目的纸浆模塑热压成型模具的加热板工作面温度分布状况,是影响产品成型质量的主要原因之一。对原加热板的孔道结构进行优化,以得到工作面温度分布更均匀的优化方案。方法用Ansys软件进行温度场仿真分析,研究了加热板孔道结构对加热板工作面温度分布的影响。结果分析结果表明:原加热板工作面最大温差达到了11.6℃,优化后加热板工作面最大温差为2.2℃,优化后较优化前加热板工作面最大温差降低了81%。结论研究为纸浆模塑加热板孔道结构提供了可行的优化方案。     

An accelerator-based epithermal neutron beam design for BNCT and dosimetric evaluation using a voxel head phantom

The beam shaping assembly design has been investigated in order to improve the epithermal neutron beam for accelerator-based boron neutron capture therapy in intensity and quality, and dosimetric evaluation for the beams has been performed using both mathematical and voxel head phantoms with MCNP runs. The neutron source was assumed to be produced from a conventional 2.5 MeV proton accelerator with a thick (7)Li target. The results indicate that it is possible to enhance epithermal neutron flux remarkably as well as to embody a good spectrum shaping to epithermal neutrons only with the proper combination of moderator and reflector. It is also found that a larger number of thermal neutrons can reach deeply into the brain and, therefore, can reduce considerably the treatment time for brain tumours. Consequently, the epithermal neutron beams designed in this study can treat more effectively deep-seated brain tumours.     

GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities.In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]).The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10¹¹ ncm²/s.This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted.The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line (http://www.info.cern.ch/asd/geant4/geant4.html[4]).To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al₂O₃) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al₂O₃) filters, respectively.To get a good cadmium ratio, GEANT 4 simulations were used to define the design of the moderator in the inlet of the radiation channel. A graphite block of 22 cm thickness seems to be the optimal neutron moderator.The results showed that the combination of 5 cm of bismuth with 5 cm of sapphire permits the filtration of gamma-rays, epithermal neutrons as well as fast neutrons in a considerable way without affecting the neutron thermal flux.     

Neutron response function for a detector with He counters for the 0.39–1.54 MeV neutron energy range

An experimental study was carried out on the neutron response functions of two neutron detector arrays consisting of 39 ³He proportional counters with a polyethylene moderator for monoenergetic neutrons within the 0.39–1.54 MeV neutron energy range. Experimental data on the sensitivity of neutron counting to a change in neutron energy and the influence of the thickness of polyethylene moderator were obtained. The experimental efficiency curves were compared with the calculated response functions generated by a neutron transport code.     

The filter/moderator arrangement-optimisation for the boron-neutron capture therapy (BNCT)

The paper presents results of the numerical modelling of the fission-converter-based epithermal neutron source designed for the boron neutron capture therapy (BNCT) facility to be located at the Polish research nuclear reactor MARIA at Swierk. The unique design of the fission converter has been proposed due to a specific geometrical surrounding of the reactor. The filter/moderator arrangement has been optimised to moderate fission neutrons to epithermal energies and to remove both fast neutrons and photons from the therapeutic beam. The selected filter/moderator set-up ensures both high epithermal neutron flux and suitably low level of beam contamination. Photons originating from the reactor core are almost eliminated what is the exceptional advantage of the proposed design. It yields one order of magnitude lower gamma radiation dose than the maximum allowed dose in such a type of therapeutic facility. The MCNP code has been used for the computations.     

Initial neutronic target station studies for the national spallation neutron source (NSNS)

Results found during initial NSNS target station neutronic design efforts are reported including the success of comparing neutron sources at 1 eV and moderator performance normalized to 1 eV. The usefulness of an analytic form is demonstrated. The angular dependence of the neutron current from a moderator face is presented together with the changes in neutron current with variation of moderator width, poison plate location and moderator material. The formation of an equilibrium state at low neutron energy is also discussed.     

Experimental studies of the effect of the ortho/para ratio on the neutronic performance of a liquid hydrogen moderator for a pulsed neutron source

Liquid hydrogen is a realistic cold moderator material for high-power spallation neutron sources. The neutronic performance of a hydrogen moderator depends on the ortho/para ratio of hydrogen, and thus experimental data are needed that will clarify the ortho/para ratio effects on neutronic performance. In this study, we measured the neutronic performance of a liquid hydrogen moderator at several para hydrogen concentrations.Our experiment was performed at the Hokkaido University 45 meV electron linac facility. The neutron energy spectra were measured by the time-of-flight method. Pulse shapes were measured by the Bragg scattering of a mica crystal.The neutron energy spectra change within 20%, depending on the para hydrogen concentrations. With increasing para hydrogen concentration, the pulse peak intensity increases and the pulse width becomes narrower. Furthermore, for a decoupled moderator, the pulse decay becomes faster with increasing para hydrogen concentration. From a viewpoint of the figure of merit (FOM=I/FWHM²) the para hydrogen moderator showed almost the same performance as that of the solid methane moderator, which is considered to be a high performance moderator for pulsed neutron source.     

Study of the neutron beam line shield design for JSNS

The JSNS, a spallation neutron source of J-PARC (JAERI-KEK Joint Project of the High Intensity Proton Accelerator) has 23 neutron beam lines. In the present study, a database was formulated for an optimum shielding design using the MCNP-X code. The calculations involved two steps. In the first step, the neutron distributions were created in the typical neutron beam line with a model that included the spallation neutron source target. The neutron currents evaluated flowed from the duct into the duct wall which was the boundary source for the bulk shield surrounding the beam line. In the second step, bulk-shield calculations were performed for the various shielding materials (iron, concrete, heavy concrete and so on) used and their composites up to thicknesses of 3 m. The results were compared with each other. Composite material shields of iron and such hydrogeneous materials as polyethylene or concrete were more effective. A typical design was prepared for a beam line within 25 m distance from a moderator, as a sample.     

NECTAR—A fission neutron radiography and tomography facility

NECTAR (Neutron Computerized Tomography and Radiography) is a versatile facility for radiographic and tomographic investigations as well as for neutron activation experiments using fission neutrons.The radiation sources for this facility are two plates of highly enriched uranium situated in the moderator vessel in FRM II. Thermal neutrons originating from the main fuel element of the reactor generate in these plates fast neutrons. These can escape through a horizontal beam tube without moderation. The beam can be filtered and manipulated in order to reduce the accompanying gamma radiation and to match the specific experimental tasks.A summary of the main parameters required for experimental set-up and (quantitative) data evaluation is presented. The (measured) spectra of the neutron and gamma radiations are shown along with the effect of different filters on their behavior. The neutron and gamma fluxes, dose rates, L/D-ratios, etc. and the main parameters of the actually used detection systems for neutron imaging are given, too.     

医院中子照射器性能与特性

介绍了医院中子照射器及其相关系统的设计情况,描述了医院中子照射器的性能和特性。为验证医院中子照射器的固有安全特性,进行了4.2 mk反应性释放实验。实验结果表明,反应堆功率快速上升,在229 s时,功率达到85.7 kW峰值,随后,由于燃料元件多普勒效应和慢化剂的负温度效应会非能动地把功率限制在允许的安全水平之下。满功率运行时,各个工艺房间γ剂量率小于1μSv/h,且中子剂量率为本底水平,具有低辐射特点。