首页 | 本学科首页   官方微博 | 高级检索  
相似文献
 共查询到18条相似文献,搜索用时 93 毫秒
1.
MCNP是用来计算中子、光子、电子或者中子/光子/电子耦合问题的通用蒙特卡罗粒子输运计算程序,它以其灵活、通用的特点以及强大的功能,在诸多领域得到广泛认可和应用。但是由于其使用需要较强的专业水平,因而使得其在某些方面又显出一些弱点。对MCNP程序的发展过程以及今后的发展趋势进行了讨论,同时提出了作者的观点。  相似文献   

2.
MCNP程序是由美国Los Alamos国家实验室研制的一个大型、多功能的粒子输运蒙特卡罗程序,可计算任意三维复杂几何系统内的中子、光子、电子或中子-光子-电子耦合输运问题,还可计算临界系统的多种本征值问题。MCNP程序的用户遍及全球,国内用户保守估计也在百家以上,过去主要应用在核科学领域,如今已推广到包括医学在内的许多领域。由于蒙特卡罗计算具有数据独立、循环粒度大、负载均衡的特性。因此,很适合作并行计算。虽然从MCNP-4A程序开始,MCNP程序具有了PVM并行计算功能,但并行系统的开发一直存在这样那样的问题,以致无法正常运行。由于MCNP程序有巨大的计算需求和计算量,我们每年使用MCNP程序完成的计算量就超过万CPU小时。许多问题采用串行计算,时间周期太长。因此,迫切需要研究缩短计算周期的并行程序。  相似文献   

3.
为了快速、精确地计算高能X射线照相中的散射光子分布,提出了将该过程中的粒子输运问题近似为一个有效的纯光子输运问题。针对该纯光子输运方程提出了一种适合于计算机计算的逐级迭代求解公式,并将该公式进行了离散化,然后编写成了计算机离散程序。用蒙特卡罗程序MCNP模拟得到了该程序所需要的参数和分布函数。最后用MCNP对该程序进行了检验。对于薄客体离散程序的计算结果与MCNP符合较好,但对于厚客体二者有较大偏离。目前可以把该程序应用于一些定性的计算分析。  相似文献   

4.
蒙特卡罗方法是目前最精确的剂量计算方法,但其较长的模拟时间阻碍了它在临床治疗中的应用。基于蒙特卡罗程序MCNP4c,针对一临床头部病例,探讨了记数方法、电子和光子截断能、光子产生次级电子参数ENUM对计算速度和精度的影响,给出了在保证一定精度前提下的最佳计算模式,以获得计算速度的有效提升。Monte its long simulation and photons, and investigated based Carlo method is regarded as the most accurate method for dose calculation at present, whereas time hinders its clinic application. The effects of the tally method, the cut-off energy of electrons the secondary electron number parameter ENUM on precision and speed of MCNP4c have been on a clinical case to seek for a relatively optimum calculation mode.  相似文献   

5.
超短超强激光与等离子体相互作用可以产生高能的超热电子, 利用光核反应的方法可以对这部分超热电子的温度进行诊断. 本文通过粒子输运程序(MCNP), 模拟了超热电子通过轫致辐射产生γ 光子, γ 光子再分别与63Cu, 107Ag, 12C等活化材料发生光核反应的物理模型, 并根据核素的活化截面数据, 计算了不同活化片的放射性活度, 得到了11C/62Cu, 11C/106Ag活度比与电子温度关系曲线, 采用理论模拟的方法实现了激光等离子体产生的超热电子的温度诊断. 关键词: 超热电子 轫致辐射光子 光核反应 MCNP程序  相似文献   

6.
在光子与电子沉积能的数值计算中,电子截断能的高低显著地影响计算效率与精度.关于电子截断能的选取,目前没有统一的标准.给出一种根据网格尺度、网格材料与电子位置自适应确定电子截断能的方法.自适应截断方法以电子在网格中的剩余总行程小于其到网格界面的最短距离作为截断条件.同MCNP程序自带的CUT方式相比,自适应方法的计算精度...  相似文献   

7.
基于自主研制的三维中子?光子耦合输运蒙特卡罗通用程序JMCT(J Monte Carlo Transport Code),采用连续点截面,对国际基准屏蔽VENUS?III模型开展精细建模和中子输运临界及屏蔽计算。临界计算得到系统kef、重要区域的通量及能谱。结果表明,JMCT和MCNP程序的重要区域体通量计数吻合较好,偏差均在1%以内。深穿透屏蔽计算采用外源模式,点探测器计数,JMCT计算值与实验测量值偏差在15%以内,满足屏蔽设计对误差的要求。初步验证了JMCT程序临界及屏蔽计算的可用性。  相似文献   

8.
聚变能源很可能是人类文明得以维持发展的新型能源。未来的氘氚聚变堆的结构和工程设计很大程度上依赖于以聚变中子学为基础的计算。在过去的十余年中,很多的核数据库如FENDL和JENDL的检验工作围绕ITER设计而展开。聚变中子学计算包括中子和光子的输运计算。其计算目标是提供反应率和能谱等重要的信息。一维或二维的聚变中子学解析计算能提供一定精度的结果和高效率的优化设计,但对于一个三维的聚变托卡马克反应堆来说,只有蒙特卡罗方法能提供较精确的数值模拟结果。MCNP程序是由LANL实验室发展的用于中子和光子的蒙特卡罗计算的大型程序。PVM的并行计算环境能提高为MCNP程序的运行执行效率。  相似文献   

9.
基于蒙特卡罗模拟方法,采用MCNP的多群计算程序模拟中子输运方程,并与栅元均匀化程序WIMS耦合,实现了临界-燃耗耦合计算。具体过程是:首先扩展MCNP的多群功能,将其能群扩展为69群;然后,由接口程序将WIMS程序产生69群共振、自屏宏观中子截面转化为ACE格式的多群截面;其次,将新产生的多群截面提供给MCNP,完成临界-燃耗计算;最后,利用此耦合程序进行了基准题校核计算以及实验对比。计算结果表明,此耦合程序是可靠和正确的。  相似文献   

10.
三维中子-光子输运蒙特卡罗程序MCMG发展了针对物质的碰撞机制,几何块、几何面动态可扩展, 随机数周期进一步扩大到261。可进行多群-连续截面耦合计算,多群散射展开到P5,并考虑了中子上散射,程序配备了通用和专用多群截面库。MCMG模拟取得了与MCNP程序和实验一致的结果,串行计算速度较MCNP快2~4倍,可进行上万处理器核的并行计算。  相似文献   

11.
In this paper, we have addressed the problem of the radiation transport with the Monte Carlo N particle(MCNP) code. This is a general purpose Monte Carlo tool designed to transport neutron, photon and electron in three dimensional geometries. To examine the performance of MCNP5 code in the field of external radiotherapy, we performed the modeling of an Electron Density phantom (EDP) irradiated by photons from 60Co source. The model was used to calculate the Percent Depth Dose (PDD) at different depths in an EDP. One field size for PDD has been examined. A 60Co photons source placed at 80 cm source to surface distance (SSD). The results of calculations were compared to TPS data obtained at National Institute of Oncology of Rabat.  相似文献   

12.
采用OKTAVIAN脉冲球实验对钍基熔盐堆用AMPX主库格式238群中子-48群光子耦合多群常数库进行了屏蔽基准验证,重点检验了该库中的F,Li,Be,C、Al,Si,Cr,Ni,Zr,Co,Cu,Mn,Mo,Nb,Ti,W,Pb同位素/元素的数据。采用SCALE 5.1程序系统中的XSDRN-PM程序进行一维屏蔽问题计算,将计算结果与实验测量数据及MCNP程序计算结果进行比较,发现中子泄漏谱的符合程度较好,而光子泄漏谱检验中发现大多数核素都出现了不同程度的高估。通过对GENDF格式到AMPX格式的转换程序MILER-4进行修正,解决了这一问题。通过对多群常数库的屏蔽基准验证,进一步证明了该库的可靠性。OKTAVIAN pulsed sphere experiment was used for Shielding Benchmarks of the AMPX formatted multi-group (238n-48γ) coupled neutron-gamma cross-section library for Thorium Molten Salt Reactor, of which the following isotopes/elements were checked-F, Li, Be, C, Al, Si, Cr, Ni, Zr, Co, Cu, Mn, Mo, Nb, Ti, W. One dimension shielding problem was calculated using XSDRN-PM program of SCALE 5.1 code system and results were compared with experiment results and MCNP calculated results, which shows that neutron leakage spectra agree well. Calculated results of photon leakage spectra of most facilities compared with MCNP results and experiment data are over-rated. MILER-4 code which is used for converting GENDF files produced by NJOY to the AMPX master library format is revised to solve this problem. The shielding benchmark verifications confirm the reliability of this new library.  相似文献   

13.
主要关于上海同步辐射装置(SSRF)储存环电子引发产生的韧致辐射和中子辐射的研究. 中子和光子经多种组合材 料(厚度在5cm~115cm之间)屏蔽后的剂量特征由蒙特卡罗代码MCNP和EGSnrc估算得到; 蒙特卡罗计算表明, 单一的材料如铅, 铁和聚乙烯对高能中子是无效的生物屏蔽材料, 而组合材料如铅或者铁加聚乙烯和铅或者铁加混凝土被认为是屏蔽高能中子很好的组合材料. 铅铁等高Z材料加点包含有氢的低Z材料如聚乙烯是同时屏蔽高能中子和韧致辐射的一种比较好的组合材料选择.  相似文献   

14.
为研究新型复合屏蔽材料的最佳厚度与各种成分最佳配比, 用MCNP计算了中子、 γ射线在稀土 高分子与重金属复合材料中的通量。 对中子、 γ射线在屏蔽体中变化规律进行了深入探索, 同传统复合屏蔽材料的屏蔽性能进行了对比。 结果表明, 中子和γ射线通过屏蔽体时, 其强度遵循指数衰减规律。 新型屏蔽材料对中子的屏蔽效果均优于铅硼聚乙烯, 对γ射线的屏蔽效果均劣于W Ni合金, 且并非稀土含量越高, 材料对中子辐射屏蔽能力越强。 A series of shielding analyses have been performed to estimate the material composition and optimum thickness required for a new radiation shield with various rare earth doped polymer and heavy metal mixtures. The neutron and γ photon fluxes have been calculated by Monte Carlo N Particle(MCNP) transport code. The results indicate that the relative fluxes of γ photon and neutron in both traditional and new composite materials follow an exponential decay rule with the distance of penetration. It can be seen that the composite material consisting of rare earth doped polymer and heavy metal has stronger neutron shielding performance than lead boron polyethylene, but weaker γ shielding effectiveness than W Ni alloy. It is also found that materials with more components of rare earth elements don’t always provide better neutron shielding performance.  相似文献   

15.
Monte Carlo (MC) codes for neutron transport calculations such as MCNP, MCNPX, FLUKA, PHITS, and GEANT4, crucially rely on cross sections that describe the interaction of neutrons with nuclei. For neutron energies below 20 MeV, evaluated cross sections are available that are validated against experimental data. In contrast, for high energies (above 20 MeV) experimental data are scarce and, for this reason, every neutron transport code is based on theoretical nuclear models to describe interactions of neutrons with nuclei in matter. Here we report on the calculation of a complete set of response functions for a Bonner spheres spectrometer (BSS), by means of GEANT4 using the Bertini and Binary Intranuclear Cascade (INC) models for energies above 20 MeV. The recent results were compared with those calculated by MCNP/LAHET and MCNP/HADRON MC codes. It turns out that, whatever code used, the response functions were rather similar for neutron energies below 20 MeV, for all 16 detector/moderator combinations of the considered BSS system. For higher energies, however, differences of more than a factor of 2 were observed, depending on neutron energy, detector/moderator combination, MC code, and nuclear model used. These differences are discussed in terms of neutron fluence rates measured at the environmental research station (UFS), “Schneefernerhaus”, (Zugspitze mountain, Germany, 2650 m a.s.l.) for energies below 0.4 eV (thermal neutrons), between 0.4 eV and 100 keV (epithermal neutrons), between 100 keV and 20 MeV (evaporation neutrons), and above 20 MeV (cascade neutrons). In terms of total neutron fluence rates, relative differences of up to 4% were obtained when compared to the standard MCNP/LAHET results, while in terms of total ambient dose equivalent, relative differences of up to 8% were obtained. Both the GEANT4 Binary INC and Bertini INC gave somewhat larger fluence and dose rates in the epithermal region. Most relevant for dose, however, those response functions calculated with the GEANT4 Bertini INC model provided about 18% less neutrons in the cascade region, which led to a roughly 13% smaller contribution of these neutrons to ambient dose equivalent. It is concluded that doses from secondary neutrons from cosmic radiation as deduced from BSS measurements are uncertain by about 10%, simply because of some differences in nuclear models used by various neutron transport codes.  相似文献   

16.
MC程序并行设计及提高加速比措施   总被引:4,自引:0,他引:4  
MC程序的并行设计涉及算法及模块划分,它直接关系到并行加速效率的高低,中子-γ耦合输运蒙特卡罗程序MCNP经过行改造,实现了PVM和MPI两种系统下的并行化,由于作了模块化设计,并行加速效率极佳,PVM版和MPI版并行程序在多个处理器下的加速比均呈线性增长,相比PVM,MPI的适应性列强,多数情况下其效率高于OPVM,并行MCNP程序的计算结果可靠,MPI并行程序在16、32和64个处理器上的并行效率分别达到99%、97%和89%。  相似文献   

17.
瞬发中子密度衰减法计算中子代时间   总被引:2,自引:1,他引:1       下载免费PDF全文
采用蒙特卡罗程序MCNP计算了西安脉冲堆中子代时间。使用MCNP程序模拟了反应堆瞬发中子通量密度衰减,基于忽略缓发中子项的点堆动力学方程计算出中子代时间。在微次临界下,研究了次临界度、源的分布、计数区域等对西安脉冲堆中子代时间计算结果的影响。计算分析表明:采用瞬发中子密度衰减法计算中子代时间时,微次临界度、源分布、计数区域等对计算结果影响都很小;误差产生的主要原因是忽略缓发中子项的点堆动力学方程并不能较好地反应瞬发中子通量密度的衰减规律。  相似文献   

18.
Monte Carlo N-particle (MCNP) code has been used to simulate the transport of gamma photon rays of different energies (22, 31, 59.5 and 81 keV) to estimate the iron content in solutions. In this study, MCNP simulation results are compared with experiment and XCOM theoretical data. The simulation shows that the obtained results are in good agreement with experimental data, and better than the theoretical XCOM values. The study indicates that MCNP simulation is an excellent tool to estimate the iron concentration in the blood samples. The MCNP code can also be utilized to estimate other trace elements in the blood samples.  相似文献   

设为首页 | 免责声明 | 关于勤云 | 加入收藏

Copyright©北京勤云科技发展有限公司  京ICP备09084417号