Application and Development Overview MCNP in Nuclear Medicine

随着计算机技术的快速发展,蒙特卡罗方法在核医学中的应用越来越广泛.概述了MCNP的基本信息,介绍了其发展情况.以MCNP5为基础,阐述了其在核医学方面应用,介绍了近年来MCNP医学物理几何数据库发展情况,同时对MCNP6的一些新特性进行了研究.     

Research Progress of Radiation-proof Material

综述了X射线、γ射线和中子辐射屏蔽材料的研究现状,其中在稀土高分子防辐射材料的设计与制备方面有所侧重,并且对纳米技术的应用、屏蔽材料的优化设计等进行了简单分析、介绍,指出了未来防辐射材料研究的可能的几个主要发展方向：纳米技术及稀土材料在防辐射材料方面的应用及研究;综合辐射屏蔽材料的设计与制备,使材料兼具质轻、无毒、体积小、屏蔽范围广、屏蔽性能持久等性能;屏蔽材料物理性能优化,以提升材料拉伸强度、硬度、耐腐蚀性等;屏蔽材料的优化设计方法、遗传算法、MCNP程序、梯度材料设计等的研究与应用;以上几个方向的交叉研究与应用。     

防辐射材料的研究进展

综述了X射线、γ射线和中子辐射屏蔽材料的研究现状,其中在稀土高分子防辐射材料的设计与制备方面有所侧重,并且对纳米技术的应用、屏蔽材料的优化设计等进行了简单分析、介绍,指出了未来防辐射材料研究的可能的几个主要发展方向:纳米技术及稀土材料在防辐射材料方面的应用及研究;综合辐射屏蔽材料的设计与制备,使材料兼具质轻、无毒、体积小、屏蔽范围广、屏蔽性能持久等性能;屏蔽材料物理性能优化,以提升材料拉伸强度、硬度、耐腐蚀性等;屏蔽材料的优化设计方法、遗传算法、MCNP程序、梯度材料设计等的研究与应用;以上几个方向的交叉研究与应用。     

稀土光功能材料发展现状及趋势

稀土光功能材料作为照明、显示和信息探测等领域的支撑材料之一,已在室内照明、液晶显示、核医学成像、辐射探测、激光工业和国防安全等领域得到广泛应用。介绍了中国稀土光功能材料的技术和产业发展现状,指出了绿色照明、高品质显示和高敏信息探测等应用新趋势,重点分析了超高色域显示和全光谱健康照明荧光粉、稀土激光和闪烁晶体等稀土光功能材料发展新动态,提出了稀土光功能材料未来发展建议。     

间规聚苯乙烯的改性研究

间规聚苯乙烯(sPS)问世以来,这种新的热塑性聚合物在不同领域的应用得到广泛关注。介绍了sPS的结构与性能,综述了其发展、改性及应用情况。     

光化学烧孔及其在频域光存储中的应用

光化学光谱烧孔是近年来发展起来的一种高新技术,在超高密度光学存储中有着光明的应用前景。本文着重从材料科学的角度介绍了其基本原理、材料特性及其在频域光存储技术中应用的最新进展情况。     

高效液相色谱-核磁共振联用技术最新进展

介绍了ＨＰＬＣ－ＮＭＲ技术的最新发展状况,讨论了ＨＰＬＣ－ＮＭＲ技术在应用中面临的问题和解决方法,评述了ＨＰＬＣ－ＮＭＲ两种主要操作方式：连续流和驻流方式,并介绍了ＨＰＬＣ－ＮＭＲ技术的应用情况,最后对其今后的发展方向作了预测。     

液相色谱–核磁共振联用技术研究进展

介绍了液相色谱–核磁共振(LC–NMR)联用技术发展状况。讨论了LC–NMR技术应用中面临的问题和解决方法,评述了LC–NMR和LC–SPE–NMR两种工作模式。介绍了LC–NMR在天然产物分析、生物代谢、异构体的鉴定和多聚物分析领域的应用情况,对其发展动态进行了展望。     

富氧膜技术及其应用

富氧膜技术是70年代末期发展起来的高效节能新技术．本文对其发展历史、国内外概况以及应用情况做了简单介绍。     

锗酸铋闪烁晶体的研究综述

锗酸铋(BGO)晶体是一种性能良好的闪烁晶体,在高能物理、空间科学、核医学等方面有着广泛的应用.作者综述了BGO晶体的发展历史、晶体的结构、晶体的生长及性能,从光谱特性、闪烁性能、辐照损伤和晶体缺陷三方面详细讨论了近年来BGO晶体的研究进展,分析了BGO晶体的发展前景.     

Two-surface Monte Carlo with basin hopping: quantum mechanical trajectory and multiple stationary points of water cluster

The efficiency of the two-surface monte carlo (TSMC) method depends on the closeness of the actual potential and the biasing potential used to propagate the system of interest. In this work, it is shown that by combining the basin hopping method with TSMC, the efficiency of the method can be increased by several folds. TSMC with basin hopping is used to generate quantum mechanical trajectory and large number of stationary points of water clusters.     

Computational Investigation on the Surface Electronic Density, Morphology and Detonation Property of 1,1-Diamino-2,2-dinitroethylene （FOX-7） Crystal

The present study constructed and optimized FOX-7 crystal using a novel technique including grand canonical monte carlo (GCMC), density functional theory (DFT) and molecular dynamics (MD) methods. Therein, the crystal density, atomic and electronic actions were considered. The results showed that the 1.96 g?cm-3 FOX-7 crystal has the highest stability and detonation properties, such as the total crystal energy, surface electronic density, friction sensitivity, detonation pressure, and so on. These results are close to the experimental data.     

Intermediate water structures in solution of Nα-acetyl-N-methylphenylalaninamide

Intermediate water structures in a solution of N^α-acetyl-N-methylphenylalaninamide[Phe(amide)₂] were investigated by the energy minimization method. The results show that ninety-eight water molecules hydrate Phe(amide)₂ and several kinds of cyclic structures are observed. The distribution of water molecules around Phe(amide)₂ agrees with the results of molecular dynamics and monte carlo studies. The distribution of cyclic structures shows that the six-membered ones are distributed mainly at the outside of the first hydration shell.     

Monte Carlo modelling of TRIGA research reactor

The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN–LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file “up259”. The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.     

Parameter optimization and experiment verification for a beta radioluminescence nuclear battery

In a beta radioluminescence nuclear battery, the beta energy is converted to light with the phosphor material, and then to electricity via photovoltaic cells. A method to optimize the thickness of phosphor layer is established in this study; the match between the luminescence spectrum and the photovoltaic cell is analyzed. The optimal parameters and output performance of the nuclear battery based on a sandwich-structure ¹⁴⁷Pm/ZnS:Cu/photovoltaic cell are determined with the MCNP, transport theory of light, and detailed balance limit of efficiency. The battery prototypes are fabricated and tested, and the experimental optimal thickness matches that of the theoretical result well.     

Electrochemical and spectroelectrochemical properties of neptunium(VI) ions in nitric acid solution

A benchmark study was carried out to verify whether MCNP is useful in the design stage of a PGNAA facility for large samples up to 1 m length and 0.15 m diameter, using a 2.54 cm diameter thermal neutron beam. For this facility neutron self-shielding and gamma-attenuation correction methods have to be developed. The relative spatial neutron-density distributions within three samples with different macroscopic scattering and absorption cross sections were studied in a comparison between an MCNP simulation and an irradiation experiment using copper wires as neutron monitors. The neutron density in the sample was within statistical agreement between experiment and simulation. Typically the relative spatial neutron-density distributions agreed to within 1%. Therefore, MCNP can be used in design studies for the development of a large sample PGNAA facility as specified. This revised version was published online in July 2006 with corrections to the Cover Date.     

Software for X- and gamma-ray spectrometry

Summary A computational approach to the true-coincidence summing correction factor evaluating was developed on the basis of the extended version of the MCNP - a general Monte Carlo N-particle transport code. A specially developed utility program generates the MCNP input on the basis of a routinely updated Evaluated Nuclear Structure Data File (ENSDF) library. Necessary information is automatically added to allow accurate simulating of the emission of correlated particles accompanying the decay of a particular radionuclide, including emission of annihilation quanta, K- and L- X-rays, β-particles and conversion electrons. Gamma-ray angular correlations as well as lifetimes of the nuclear excited states are also taken into account. The approach is applicable to correction factor evaluation for ordinary single- and multi-detector spectrometers as well as for Compton suppression systems. The paper describes the developed computational scheme as well as presents the results of its preliminary testing for the case of both point and volumetric sources.     

Validity of two simple rescaling methods for electron/beta dose point kernels in heterogeneous source–target geometry

Dose calculations around electron-emitting metallic spherical sources were performed up to the X₉₀ distance of each electron energy ranging from 0.5 to 3.0 MeV using the MCNP 4C Monte Carlo code and the dose point kernel (DPK) method with the DPKs rescaled using the linear range ratio and physical density ratio, respectively. The results show that the discrepancy between the MCNP and DPK results increases with the atomic number of the source (i.e., heterogeneity in source–target geometry), regardless of the rescaling method used. The observed discrepancies between the MCNP and DPK results were up to 100% for extreme cases such as a platinum source immersed in water.     

Correction for neutron self-shielding in large-sample prompt-gamma neutron activation analysis

When a sample is analysed with neutron activation analysis (NAA) neutron self-shielding and gamma self-absorption affect the accuracy. Both effects become even more important when the mass of a sample analysed is changed from small (say, 1 g) to large (say 30 kg). Therefore, corrections have to be carried out. In this article only the correction method for neutron self-shielding is considered for a thermal neutron beam irradiating large homogeneous samples for prompt-gamma NAA (PGNAA). The correction method depends on the macroscopic scattering and absorption cross sections of the sample. To avoid doing experiments with samples with different macroscopic scattering and absorption cross sections, the Monte Carlo model MCNP is applied in the development of the correction method. The computational development of the method to determine these cross sections through flux monitoring outside the sample is described.