共查询到20条相似文献,搜索用时 250 毫秒
1.
2.
3.
为分析含孔隙结构地层的γ射线响应特性,编写了一种Monte Carlo方法模拟γ射线散射的程序,使用该程序除了能建立一般的几何模型外,还能在数字岩心的基础上进行模拟。利用该程序对γ光子在NaI晶体中的脉冲沉积谱做了模拟,并与MCNP4C相应的计算结果进行了对比,对比显示,该程序能提供正确、有效的模拟手段。在正确性验证的基础上,以数字岩心构建地层,对不同孔隙度的砂岩做了密度测井的能谱模拟,得到了一些有益的结果。 相似文献
4.
5.
计算机模拟计算在各种类型加速器的设计中起着很重要的作用。例如,计算机模拟很快就可以确定加速器各部件的最佳工作参数和所需要的光学条件,使一台加速器复杂的束流输运系统设计变成简单的工作。TRANSPORT程序是当今世界上关于束流输运计算的最优秀程序之一。但是,该程序不能进行静电加速器离子光学系统的计算,也不能作直线加速器的粒子动力学计算。为此从北京大学引进了LEADS程序,它不仅具有TRANSPORT程序的优点,还克服了其不足。对含有聚束(脉冲束流)系统的加速器粒子动力学进行模拟是LEADS程序的新特色。实验中用LEADS程序模拟了200keV直流/脉冲中子发生器束流输运过程,得到了横向相图、纵向相图和束包络图。还用LMOV程序计算和验证了聚束系统,束流脉冲波形图和实验结果比较接近。 相似文献
6.
7.
聚束系统设计与计算程序 总被引:2,自引:0,他引:2
LMOVE程序是一个模拟粒子纵向运动的FORTRAN语言程序。它可以计算由三电极双间隙聚束器(或这种聚束器的组合)、单狭缝谐波聚束器、静电加速管和自由飘浮空间组成的束流脉冲化系统。程序采用大量粒子模拟束团在纵向相空间内的运动,并用Monte-Carlo方法随机产生相空间内各代表点的初始条件。因此,较为适合模拟粒子在配有聚束器件的静电加速器中的纵向运动。本程序不考虑粒子的空间电荷效应。 相似文献
8.
9.
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。 相似文献
10.
在非能动安全壳冷却系统(PCS)设计基准事故的排热过程中,安全壳内壁面蒸汽冷凝现象和安全壳外壁面水膜蒸发现象是两种非常关键的排热途径。本文应用GOTHIC8.0程序模拟了安全壳内壁面蒸汽冷凝和安全壳外壁面水膜蒸发传热过程,并通过蒸汽冷凝试验和水膜蒸发试验数据,对GOTHIC程序的模拟结果进行了分析和评价。研究结果表明:GOTHIC程序的蒸汽冷凝模型可较好地模拟蒸汽冷凝传热现象;水膜蒸发模型明显低估了水膜蒸发换热量,这对设计基准事故安全壳完整性分析是非常保守的,建议对GOTHIC程序进行适当开发,更好地模拟水膜蒸发换热过程。 相似文献
11.
The article provides an overview of the reactor dynamics code DYN3D. The code comprises various 3D neutron kinetics solvers, a thermal-hydraulics reactor core model and a thermo-mechanical fuel rod model. The implemented models and methods and the capabilities and features of the code are described. Latest developments of models and methods are delineated. An overview on the status of verification and validation is given. Code applications for selected safety analyses are described. Furthermore, multi-physics code couplings to thermal-hydraulic system codes, CFD and sub-channel codes as well as to the fuel performance code TRANSURANUS are outlined. Developments for innovative reactor concepts, in particular Molten Salt Reactor, High Temperature Gas-cooled Reactor and Sodium Fast Reactor are delineated. The management of code maintenance is briefly described. An outlook on further code development is given. 相似文献
12.
本文基于Cinder90燃耗数据库开发了燃耗求解程序MCRAM,并耦合MCNP程序对重要的锕系核素和裂变产物核素的反应截面进行了修正。以OECD/NEA乏燃料成分基准数据库中的Takahama-3压水堆燃料组件为基准题,对MCRAM程序的计算结果进行了验证,并与其他程序的计算结果进行了比较。结果表明,MCRAM程序对重要裂变产物和主要锕系核素的计算结果相对偏差小于5%,计算精度与ORIGEN2程序的相当。与此同时,同一例题的计算效率MCRAM较之MCNTRANS程序提高了近200倍。 相似文献
13.
14.
A multigroup general purpose Monte Carlo code GMVP has been developed. The vectorization algorithm is based on a stack-driven zone selection method. GMVP can treat repeated rectangular and hexagonal lattices together with combinatorial geometry which is quite useful to achieve a high gain by vectorization. The performance of the code was evaluated by solving various types of problems. In addition, the continuous energy code is under development and the performance is compare with conventional codes. The code was installed on other four different supercomputers to investigate portability and computer dependence of code performance. 相似文献
15.
介绍了三维带控制棒计算反应性温度系数的计算方法,对于200MW供热堆,在不同棒位下作了三维反应性温度系数的计算,结果表明,慢化剂温度系数的大小和控制棒插入的深度有一定关系。以运行工况临界棒位下的慢化剂温度系数为参考,对二维计算结果作了分析,结果表明第二维无控制棒计算是保守的近似。 相似文献
16.
Ashutosh Tiwari Brian AllisonJ.K. Hohorst R.J. WagnerChris Allison 《Fusion Engineering and Design》2012,87(2):156-160
RELAP/SCDAPSIM Mod 4.0 code was developed by Innovative System Software (ISS) for the analysis of nuclear power plants (NPPs) cooled by light water and heavy water. Later on the code was expanded to analyze the NPPs cooled by liquid metal, in this sequence: lead bismuth eutectic mixture, liquid sodium and lead lithium eutectic mixture (LLE) are inserted in the code. This paper focuses on the insertion of liquid LLE as a coolant for NPPs in the RELAP/SCDAPSIM Mod 4.0 code. Evaluation of the code was made for a simple pipe problem connected with heat structures having liquid LLE as a coolant in it. The code is predicting well all the thermodynamic and transport properties of LLE. 相似文献
17.
铀氢锆堆物理计算及燃料管理软件包 总被引:3,自引:1,他引:2
建立了一套铀氢锆堆物计算软件包,首先考虑氢化锆中的热化特殊性,按WMS格式制作 了氢化锆 氢的69群群常数并入WIMS-D/4数据库中,形成了WIMS-N1库和WIMS-N2库;应用WIMS-N2库和国际通用的WIMS-D/4程序包计算了铀氢锆堆各类栅元的群常数,应用差分程序CITATION和六角形节块和SIXTUS进行扩散计算,同时在SIXTUS-2程序的基础上编制了燃料管理程序和XPR-ICF 相似文献
18.
M.B. Carver A. Tahir J.C. Kiteley A.O. Banas D.S. Rowe W.I. Midvidy 《Nuclear Engineering and Design》1990,122(1-3)
ASSERT-4 is a subchannel code based on the non-equilibrium equations of two-fluid flow. The paper briefly describes the equations and constitutive models used in the code, and reviews a number of validation exercises in which code results were compared to measurements in vertical and horizontal two-phase flows. 相似文献
19.
20.
N. Catsaros B. Gaveau M. Jaekel J. Maillard G. Maurel P. Savva J. Silva M. Varvayanni Th. Zisis 《Annals of Nuclear Energy》2009,36(11-12):1689-1693
In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal–hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the “Accelerator part” of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the “Reactor part” of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes. 相似文献