Penumbral imaging and numerical evaluation of large area source neutron imaging system

The fusion neutron penumbral imaging system Monte Carlo model was established. The transfer functions of the two discrete units in the neutron source were obtained in two situations: Imaging in geometrical near-optical and real situation. The spatial resolutions of the imaging system in two situations were evaluated and compared. The penumbral images of four units in the source were obtained by means of 2-dimensional (2D) convolution and Monte Carlo simulation. The penumbral images were reconstructed with the same method of filter. The same results were confirmed. The encoding essence of penumbral imaging was revealed. With MCNP(Monte Carlo N-particle) simulation, the neutron penumbral images of the large area source (200 μm×200 μm) on scintillation fiber array were obtained. The improved Wiener filter method was used to reconstruct the penumbral image and the source image was obtained. The results agree with the preset neutron source image. The feasibility of the neutron imaging system was verified. Supported by the NSAF Joint Fund set up by NSFC and CAEP (Grant No. 10576022)     

MCNP Output Data Analysis with ROOT (MODAR)

MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time-energy data issued by MCNP simulations of neutron inspection devices using the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a fast and user-friendly way. MODAR allows to take into account the detection system time resolution (which is not possible with MCNP) as well as detectors energy response function and counting statistics in a straightforward way.

Program summary

Program title: MODARCatalogue identifier: AEGA_v1_0Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGA_v1_0.htmlProgram obtainable from: CPC Program Library, Queen's University, Belfast, N. IrelandLicensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.htmlNo. of lines in distributed program, including test data, etc.: 155 373No. of bytes in distributed program, including test data, etc.: 14 815 461Distribution format: tar.gzProgramming language: C++Computer: Most Unix workstations and PCOperating system: Most Unix systems, Linux and windows, provided the ROOT package has been installed. Examples where tested under Suse Linux and Windows XP.RAM: Depends on the size of the MCNP output file. The example presented in the article, which involves three two-dimensional 139×740 bins histograms, allocates about 60 MB. These data are running under ROOT and include consumption by ROOT itself.Classification: 17.6External routines: ROOT version 5.24.00 (http://root.cern.ch/drupal/)Nature of problem: The output of an MCNP simulation is an ASCII file. The data processing is usually performed by copying and pasting the relevant parts of the ASCII file into Microsoft Excel. Such an approach is satisfactory when the quantity of data is small but is not efficient when the size of the simulated data is large, for example when time-energy correlations are studied in detail such as in problems involving the associated particle technique. In addition, since the finite time resolution of the simulated detector cannot be modeled with MCNP, systems in which time-energy correlation is crucial cannot be described in a satisfactory way. Finally, realistic particle energy deposit in detectors is calculated with MCNP in a two-step process involving type-5 then type-8 tallies. In the first step, the photon flux energy spectrum associated to a time region is selected and serves as a source energy distribution for the second step. Thus, several files must be manipulated before getting the result, which can be time consuming if one needs to study several time regions or different detectors performances. In the same way, modeling counting statistics obtained in a limited acquisition time requires several steps and can also be time consuming.Solution method: In order to overcome the previous limitations, the MODAR C++ code has been written to make use of CERN's ROOT data analysis software. MCNP output data are read from the MCNP output file with dedicated routines. Two-dimensional histograms are filled and can be handled efficiently within the ROOT framework. To keep a user friendly analysis tool, all processing and data display can be done by means of ROOT Graphical User Interface. Specific routines have been written to include detectors finite time resolution and energy response function as well as counting statistics in a straightforward way.Additional comments: The possibility of adding tallies has also been incorporated in MODAR in order to describe systems in which the signal from several detectors can be summed. Moreover, MODAR can be adapted to handle other problems involving two-dimensional data.Running time: The CPU time needed to smear a two-dimensional histogram depends on the size of the histogram. In the presented example, the time-energy smearing of one of the 139×740 two-dimensional histograms takes 3 minutes with a DELL computer equipped with INTEL Core 2.     

14MeV中子防护体的设计

在用D—T中子发生器分析煤质时,为了设计14MeV中子防护体,本文先合理简化了实际模型,然后用MC-NP程序对其进行模拟,最后用实验结果修正模拟结果．实验和模拟结果显示,当聚乙烯厚度超过300mm时,地面的中子注量率就达到了辐射防护的要求．为了安全起见,在实际设计中用500mm的聚乙烯防护14MeV中子．     

蒙特卡罗粒子输运计算自动建模程序系统的研究与发展

在广泛调研和分析现有几何建模方法特点的基础上研发了具有可视化用户界面的自动建模程序系统MCAM.它可以实现多种商用软件CAD模型与MCNP模型之间的相互转换,且提供了模型建立、预处理、属性分析等基本功能和计算结果可视化及基于医学映像建模接口等扩展功能.全面系统地介绍了MCAM的设计思想与原理、总体结构、主要功能和国际合作协议框架下的应用测试等情况.实践表明,它是一个实用的MCNP计算辅助工具和核设计与核分析质量保证工具.     

钨-铜合金影锥屏蔽体屏蔽性能的MCNP程序计算

利用MCNP程序对影锥屏蔽体的屏蔽性能进行计算和深入分析。结果表明：影锥屏蔽体对于周围及样品造成的散射中子本底影响低于1.4%。中子穿透影锥屏蔽体而产生的γ射线泄漏率为10^-16～10^-14数量级，对于中子散射微分截面的实验测量，其影响可以忽略不计。W-Cu合金影锥屏蔽体的设计模型符合设计标准，就飞行距离为4～10 m的范围而言，影锥屏蔽体可使源中子注量衰减10^-7，屏蔽效果显著。     

蒙特卡罗方法在中子活化在线分析系统设计中的应用

选取重水、石墨、聚乙烯等6种慢化材料，利用MCNP程序对不同的慢化材料进行模拟计算分析。计算结果表明，中子活化在线分析系统的最优化慢化材料为聚乙烯。实验测定了以聚乙烯为慢化材料的中子活化分析系统的热中子注量率随源到引出孔之间的距离以及探测器处于不同位置时的分布关系，为下一步进行中子活化在线分析研究提供了依据。     

MCNP程序在ADS屏蔽计算中的应用

利用基于蒙特卡罗方法的MCNP程序计算了加速器驱动的次临界系统(ADS)中质子束管内的中子注量率分布以及通过质子束管顶端面和其它外表面逸出的中子注量率，得出了一些对ADS设计有意义的结论。     

使用MCNP程序调整和优化启明星实验装置设计方案

用蒙特卡罗程序(MCNP)对验证ADS系统的启明星实验装置的设计方案进行了有效增殖因数(Keff)计算,并对与Keff密切相关的热区燃料元件栅距和热区厚度进行了最优参数的计算。结果表明,启明星实验装置的Keff能够达到设计的目标。     

Advanced Burnup Calculation Code System in a Subcritical State with Continuous-Energy Monte Carlo Code for Fusion-Fission Hybrid Reactor

The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before.     

Computation of the neutron multiplicity moments for research reactor fuels using MCNP6 and SOURCES4c

Fissile material detection and quantification are often necessary for safeguards, nuclear security, and fuel management. Nondestructive assay, neutron, and gamma measurements are reliable means, which can facilitate the detection and estimation of the mass of fissile materials in a broad range of material matrix. Various flavours of neutron measurement are routinely used by facilities (like nuclear reactors, enrichment, and fabrication plants) to quantify fissile material mass and inventory lists. The Monte Carlo code, MCNP6, is used to model several neutron multiplicity measurements. A simulation scenario is set up in MCNP6 using the JCC71 neutron slab counter to obtain the multiplicity moments for fresh and irradiated fuel assemblies from the UMass Lowell Research Reactor (UMLRR) and Worcester Polytechnic Research Reactor (WPIRR). An MCNP6 burnup is initially performed on the fuel types under study to generate used fuel isotopic. The fresh and or used fuel isotopic is then used to produce independent SOURCES4c input tape1 files. SOURCES4c is used to generate (α, n), spontaneous fission spectrum, and the associated neutron emission rates necessary for the various fixed fuel source definitions in MCNP6 calculations. Under the comprehensive safeguards agreements, the International Atomic Energy Agency has the right and obligation to verify that no nuclear material is diverted from peaceful use to nuclear weapons or other nuclear explosive devices. Research reactors are required to be safeguarded facilities under the comprehensive safeguards. Several research efforts have studied the various flavours of neutron measurement for commercial power reactor operating at high power and long burnups; however, not nearly as many studies have been performed with neutron measurements for research reactors operating at relatively lower power and have significantly lower burnup. This work looks to establish the relevant isotopes to overall neutron source rate as well as the process involved in performing a typical neutron multiplicity measurement simulation for a research reactor fuel. The results demonstrate that the single and double moments for Worcester Polytechnic Institute (WPI) and UMLRR fuels can be measured reliably using two JCC71 slab detectors. The moment for the UMLRR and WPIRR fuel (in both fresh and used states) was estimated with a relative error below 0.031 for singles and 0.081 for doubles. The two fresh fuel types cannot be differentiated from each other on the sole basis of neutron analysis. However, fresh and irradiated fuel can be distinguished based on neutron multiplicity measurements.