三维中子-光子输运的蒙特卡罗程序MCMG

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。     

蒙特卡罗粒子输运计算自动建模程序MCAM在ITER核分析建模中的应用

国际热核试验堆（ITER）核分析的主要计算工具是三维蒙特卡罗输运程序MCNP。MCAM（MCNP Auto-Modeling system）作为MCNP自动建模与可视化软件,其主要功能是CAD模型与MCNP计算模型之间的数据交互。一方面MCAM可将通用格式的CAD模型转换成MCNP计算模型,另一方面作为一个可视化工具,它可以以CAD模型的方式显示MCNP计算模型中几何及材料等相关信息。主要介绍利用MCAM对ITER三维MCNP模型的改进,主要包括:包层模块的重建和内包层几何细化;模型环向角度从20°到40°的扩展。In order to conduct nuclear analyses on neutronics issues for ITER （International Thermonuclear Experimental Reactor）, a standard three-dimensional model of the ITER reactor is being developed. The complex nuclearanalyses are conducted by MCNP/4C in three dimensions. MCAM （MCNP Auto-Modeling system） , as an implementation of the interface code between modern CAD system and MCNP, is a modeling and visualization tool which can convert a CAD model to neutronics model for MCNP and vice versa. This paper presents the application of MCAM to modify ITER 3-D neutronics model, which include blanket segmentation update, incorporation of fine structures of inboard blanket and the model extension from 20° to 40° in toroidal direction     

中子屏蔽虚拟仿真实验系统的设计与开发

结合兰州大学核学科特色和核工程类专业人才培养方案,从以学习者为中心的角度出发,设计开发了基于LabVIEW和MCNP的中子屏蔽虚拟仿真实验系统.系统采用友好的人机交互界面,通过模拟中子与物质的相互作用过程,可以实现不同种类、不同厚度的中子慢化/屏蔽材料和材料组合对常用中子源辐射场屏蔽效果的模拟仿真以及实验结果的可视化输...     

温度相关核截面数据库在MCNP计算中的必要性研究

MCNP程序由于其几何模拟和核数据上的优越性,现在在反应堆的研究分析中已经得到较多应用。通过基准题的计算,定量地说明MCNP通过其自带的常温（294K）下的核素截面数据库不能够对反应堆进行非常准确的计算（由于反应堆内各种材料／位置的温度不同）,而且,它也不能够计算反应堆中与温度相关的量,如反应性温度系数。选用了一个带有不同温度下核素截面数据的MCNP输入格式的数据库,使用MCNP-4C对基准题进行了计算,发现计算结果与基准值符合得非常好。这说明通过使用不同温度下的核素截面数据库,MCNP可以准确计算温度系数和增殖系数等,从而说明在反应堆设计计算中制作不同温度下的核素截面库的必要性。Due to the advantage of geometry simulation and nuclear data, the code MCNP is now widely used in the reactor analysis. Based on our calculation of the fuel temperature reactivity coefficient benchmark, it is quantificationally proved that MCNP with its own cross section library can＇ t be used to simulate the reactor accurately and to calculate the temperature reactivity coefficient. Furthermore, we use MCNP- 4C with a database that contains temperature dependent nuclear cross sections to calculate the benchmark. The results are well agreement with benchmark results. This means that, with the temperature dependent nuclear cross sections library, MCNP can calculate the temperature reactivity coefficient and reactor multiplication factor accurately. So the temperature dependent nuclear cross section library should be processed to meet the requirement of reactor calculation.     

一种分段式γ扫描系统探测效率响应函数的数值计算方法

基于SGS探测系统中放射性核废物桶、准直器及HPGe探测器的空间几何分布建立数学模型。对低、中放射性核废物桶中点源137Cs的探测效率响应函数进行了数值公式推导，然后利用蒙特卡罗工具包MCNP模拟计算了相应的效率函数。将这两种方法模拟的结果进行了对比分析发现:数值计算方法的数据普遍高于MCNP模拟的数据，最大相对误差达143.26%，误差平均值为37.15%。对数值计算方法进行修正后，最大值相对误差为17.22%，平均值为4.54%。在误差范围内，可以认为该修正方法有效。     

MCNP/MCNPX几何栅元划分方法对精确放疗剂量计算的影响研究

复杂几何模型的建立是Monte Carlo粒子输运程序MCNP／MCNPX在放疗领域广泛应用的关键与难点,发展了基于医学CT影像的MCNP／MCNPX自动建模软件,提出并实现了3种几何栅元划分的方法。根据临床实例数据,分别建立了3种MCNP几何模型。在此基础上,研究分析了3种几何栅元划分方法及重复结构描述方法对计算结果的影响,为MCNP／MCNPX在放疗中的应用提供基础。The key problem for the application of the Monte Carlo particle transport code MCNP/MCNPX in radiotherapy is the creation of complex geometrical model. To handle this problem, a software has been developed to automatically create MCNP/MCNPX geometrical model based on the CT images, and three geometric cell treatment schemes were proposed and implemented in this software. In this work, three MCNP models are created, and calculations are performed to investigate the effect of those cell treatment schemes and repeated structure technique on the calculation results.     

近垒熔合机制的动力学研究

通过考虑动力学中的壳效应, 对同位旋相关的量子分子动力学模型做了进一步改进, 并利用改进的量子分子动力学模型对近垒熔合机制做了动力学研究. 对轻系统的熔合反应, 改进的模型能够给出与实验结果一致的熔合激发函数, 同时, 对重系统的俘获截面也能给出与实验测量值基本一致的计算结果, 尤其是对近垒或垒下区域也能作出很好的描述. 作为对模型的进一步检验, 对核--核相互作用势做了静态研究, 并与亲近势的结果做了比较.     

西安脉冲堆127I嬗变计算方法与实验验证

为了开展129I的热中子嬗变的研究,在西安脉冲堆上开展了127I靶件辐照实验。以探索实验条件,对127I靶件的嬗变率进行了蒙特卡罗计算,并与实验测量值进行了比对。利用NJOY程序,以ENDF/B VII.0库为基础,制作了127I在西安脉冲堆堆芯辐照温度下的MCNP格式截面库,与MCNP自带库(ENDF/B VI.2)同温度下截面库进行了比较,在不可分辨共振区做了改进,使用新制的截面库,利用MCNP程序对ORIGEN2数据库中的127I辐射俘获截面进行了修正,结合ORIGEN2程序分析了127I靶件在西安脉冲堆实际辐照后的嬗变率和核素的变化,研究了中子能谱和辐照时间对靶件嬗变计算的影响。使用MCNPX自带的燃耗模块CINDER90对127I靶件的嬗变情况进行模拟,结果与ORIGEN2基本一致,与实验数值有2%~3%的偏差,主要原因是MCNP计算中子通量密度存在误差。     

CFETR水冷陶瓷增殖剂包层中子学分析

基于中国聚变工程实验堆(CFETR)水冷陶瓷增殖剂(WCCB)三维中子学模型,应用蒙特卡罗输运程序MCNP5和IAEA聚变评价核数据库FENDL2.1,完成了WCCB中子学性能分析。研究了在200MW、500MW、1.0GW、1.5GW聚变功率下中子壁载荷(NWL)、氚增殖率(TBR)、核热沉积以及包层材料的辐照损伤。结果显示,目前WCCB包层核分析结果满足CFETR设计要求。     

HZE beam transport in multilayered materials

A nonperturbative analytic solution of the high charge and energy (HZE) Green's function is used to implement a computer code for laboratory ion beam transport in multiple-layered materials. The code is established to operate on the Langley nuclear fragmentation model used in space engineering applications. Computational procedures are established to generate linear energy transfer (LET) distributions for a specified ion beam and target for comparison with experimental measurement. Comparison with 56Fe ion with Pb-Al and Pb-(CH2)x targets shows reasonable agreement.     

Neutron activation cross section data library

To satisfy the requirements of nuclear reaction cross sections in nuclear engineering applications and nuclear physics studies, the Neutron Activation Cross Section Data Library has been established. 818 target nuclei including unstable target or isomeric target nuclei are considered in this library. The induced neutron energy range region is between 10^-5 eV and 20 MeV. The standard ENDF-6 format is adopted, including general information, reaction cross sections, multiplicities, and so on. The recommended reaction cross sections were obtained using UNF code system and FDRR nuclear model codes or systematic analysis based on available experimental data. The full evaluated dataset containing the evaluated activation cross sections is openly available at http://doi.org/10.57760/sciencedb.j00113.00024.     

Improvement of the algorithm for the analysis of Level schemes in the GTOL code

The algorithm of the GTOL code used for constructing nuclear level schemes by the measured γ spectra is discussed, and the difficulties arising in the work with complex nuclear level schemes are pointed out. The ways to improve the algorithm are shown, and the NEWGTOL code allowing large arrays of nuclear data to be properly handled is proposed.     

Determination of the fission fragment excitation energy with allowance for neutron multiplicity

The study of the excitation energy distribution of fission fragments as a function of their mass and charge is important for revealing the nuclear fission mechanism and useful for many applications. To measure directly the excitation energy of primary fission fragments (before emission of neutrons) is a great problem. A method of obtaining these excitation energies from calculated neutron multiplicities and experimental values for differential yields of fragment pairs after emission of neutrons is considered. The Empire-II code was used to calculate neutron multiplicities as a function of various characteristics of the nuclear structure, fission process, and fission fragment deexcitation.     

Simulation of Nuclear Reactor Kinetics by the Monte Carlo Method

The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.     

Monte Carlo simulations of bremsstrahlung photon yields from thin targets with electron beams between 10 and 40 MeV

A linear accelerated electron beam is a powerful source of highly energetic photons that can be used for various nuclear applications. In this work, the bremsstrahlung photon yields from thin targets such as Ta, Nb, Cu and Al have been calculated using the FLUKA simulation code. The simulation calculations were performed for incident electrons with energies in the range of 10–40 MeV and targets with thicknesses between 4×10^?4 and 10^?3 radiation lengths.     

Au~(34+)离子双电子复合过程的理论研究

复杂结构离子的双电子复合(DR)速率系数在核聚变、极紫外光刻光源等应用研究的等离子体谱模拟中具有重要的价值.利用基于全相对论组态相互作用理论的FAC程序包,详细计算了Au~(34+)离子的双电子复合速率系数.研究分析了激发、辐射通道,组态相互作用,级联退激对DR速率系数的影响.其中,级联退激对DR速率系数的贡献必须予以考虑.对双电子复合、辐射复合以及三体复合速率系数做了比较,在温度大于1 eV范围,双电子复合都大于辐射复合以及三体复合速率系数,相应的DR过程对于等离子体离化态分布和能级布居以及光谱模拟都极为重要.对基态和第一激发态的DR速率系数进行了参数拟合,拟合值与计算值的偏差小于1.73%.研究结果将为复杂结构离子双电子复合过程的进一步研究提供参考.     

Calculations of particle and heavy ion interactions with space shielding materials using the SHIELD-HIT transport code

Spectra of secondary particles produced by nucleus–nucleus interactions and heavy-ion interactions in the extended targets of interest for space research were calculated using the Monte Carlo code SHIELD-HIT. This code simulates the interactions of hadrons and atomic nuclei of arbitrary charge and mass number (Z,A) with complex extended targets in a wide energy range, from 10 GeV/u down to 1 MeV/u, and to thermal energies in the case of neutrons. Inelastic nuclear reactions in SHIELD-HIT are simulated using the Russian models of nuclear reactions. The total reaction cross sections evaluated by these models are discussed for proton and carbon interactions with different nuclei in a wide energy range. Production of secondary neutrons and charged secondary particles from the thick targets of lead, water and PMMA irradiated by ⁴He, ¹²C and ²⁸Si ions of different energies was calculated and compared with the experimental data. The results obtained by SHIELD-HIT are in reasonable agreement with experiments and are promising for further applications in space research.     

Statistical simulation of complete events in energetic nuclear collisions

A statistical model is developed for the generation of complete multifragment events in medium-energy nuclear collisions. Based on simple geometrical considerations, the collision system is divided into a few (participant/spectator) sources that disassemble independently. The sufficiently excited sources quickly explode into pions, nucleons, and composite, possibly particle unstable, nuclei. The different final states compete according to their microcanonical weight. The less excited sources, and the unstable explosion products, deexcite by sequential light-particle emission. The model has been implemented as a Monte Carlo computer code that is sufficiently efficient to permit generation of large event samples. The analysis of such multifragment events is addressed and some illustrative applications are discussed.     

Response functions of Cs2LiYCl6: Ce scintillator to neutron and gamma radiation

Research into advanced screening technologies has become high priority in all aspects of occupational nuclear safety and environmental radiation protection. Neutrons are a fundamental part of radiation encountered in various fields of nuclear science and technology and their detection is still employing detectors with a high thermal neutron response embedded in a thermalizing medium where helium based devices have been a dominant choice in many applications. Recently, there have been newly developed sensors based on multi-elements that include ⁶Li and ³⁵Cl isotopes to detect neutrons and gamma radiation. Among these new sensors one can cite the elpasolite scintillator, known as CLYC. This sensor contains two neutron sensitive isotopes and may serve as a dual detector for gamma as well as for neutron radiation. In this paper, the response functions of this sensor have been investigated in different fields of neutron and gamma-radiation. The sensor responses have been simulated using Monte Carlo N-Particle MCNPX code and a series of experiments have been carried out to validate the simulated data. Both sets of data are presented and discussed.