点通量积分法在蒙特卡罗方法中的应用

介绍了指向概率法和点通量积分法两种方差减小技术在计算小概率的体通量、面流量中的应用,进而利用点通量积分法计算了在MCLLS方法的典型几何结构中的面流量计算结果并且与通用蒙卡程序MCNP的计算结果进行了比较。     

闪光层析成像散射修正计算研究

利用闪光照相成像系统底片信息,通过直接法对客体线性吸收系数空间分布进行重建,需能精确得到底片直穿量的信息,为解决这一问题必需对散射量进行修正。本文通过解析法推导出精确计算X光正向输运过程中未经碰撞与只经1次碰撞到达探测器处计数的理论公式,同时通过蒙特卡罗计算程序MCNP分析不同照相客体次级散射效应特性,对次级散射量进行数值修正,并由此编制快速光子正向计算程序SAT。与MCNP程序统计结果进行数值校验,结果表明,SAT程序能精确计算X光穿过照相客体到达探测器处的直穿量、散射量及总照射量计数,且计算速度远快于MCNP程序,这为闪光层析成像图像重建迭代扣除散射提供了快速精确的散射修正工具。     

MCNP程序对某坑式废物库贮存废物放射性活度验证计算

某废物库退役源项调查是废物库退役工程前期工作,为退役设计提供源项输入。根据现场辐射水平测量结果,利用点核积分原理,通过γ辐射水平与γ射线注量之间的关系,反推出废物的放射性活度。为评价计算结果,利用MCNP程序进行了验证计算。点核积分计算结果为6.4×1014Bq,MCNP验证计算结果为4.82×1014Bq,表明点核积分计算结果偏保守,满足工程设计需要。     

用于迷宫屏蔽计算的蒙特卡罗减方差方法研究

本文应用多种减方差技巧提出了强迫指向自动重要抽样（FPAIS）方法,并在MCNP5程序平台实现了该方法。采用该方法对1个多折迷宫算例进行了模拟计算,计算结果与MCNP5程序的直接模拟、DXTRAN球、点通量3种方法的结果进行了比对。基于此算例对FPAIS方法进行了引导面设置和粒子数敏感性分析。结果表明,FPAIS方法在保证一定计算精度的前提下,比其他3种方法的FOM提高2~3个量级,且该方法对引导面设置不敏感、可用性强,对于迷宫屏蔽计算是一种准确、高效的解决方案。     

用于核素在线瞬发伽马数据测量的热中子源设计

基于MCNP程序建立了西安脉冲堆热中子源设计的蒙特卡罗深穿透耦合屏蔽计算方法;采用MCNP临界源模型计算了热柱方腔前表面的中子、伽马平面源的参数,并与实验值进行了对比,给出了平面源的修正系数;基于中子、伽马等效平面源,采用新型硼铝复合材料以及铅、铋等材料,优化设计了热中子束流滤束装置,给出了热中子束流滤束装置的升级改造方案,得到热中子通量密度较原设计方案提高3倍、中子伽马通量密度比值大于10的平行热中子束,且束流外侧区域的中子、伽马本底剂量率接近0.025 mSv/h的辐射防护标准。     

用于核素在线瞬发伽马数据测量的热中子源设计

基于MCNP程序建立了西安脉冲堆热中子源设计的蒙特卡罗深穿透耦合屏蔽计算方法;采用MCNP临界源模型计算了热柱方腔前表面的中子、伽马平面源的参数,并与实验值进行了对比,给出了平面源的修正系数;基于中子、伽马等效平面源,采用新型硼铝复合材料以及铅、铋等材料,优化设计了热中子束流滤束装置,给出了热中子束流滤束装置的升级改造方案,得到热中子通量密度较原设计方案提高3倍、中子伽马通量密度比值大于10的平行热中子束,且束流外侧区域的中子、伽马本底剂量率接近0.025 mSv/h的辐射防护标准。     

基于点核法的γ辐射剂量计算软件设计

为满足核设施退役过程中需要大量剂量计算要求,利用C#4.0编程语言开发了基于点核方法的γ剂量计算程序PKShield。该程序包含传统点核程序QAD-CGPIC的大部分特性,并且提供了数据输入的图形界面,扩充了放射性源的几何类型。PKShield能够计算具有能谱的多放射性源的剂量分布,且具有较快的计算效率。为了验证PKShield计算方法的有效性和正确性,利用所开发的PKShield与MCNP5进行剂量计算结果比较,结果表明开发的PKShield程序能够正确、有效地计算γ辐射剂量。     

基于CADIS方法的蒙特卡罗深穿透屏蔽计算方差降低方法研究

正深穿透问题是指由于屏蔽层过厚、源强较弱或探测器体积较小导致粒子输运计算结果偏低的问题。针对于蒙特卡罗屏蔽计算中的深穿透问题,利用一致共轭驱动重要性抽样(CADIS)方法的相应理论,实现了蒙特卡罗软件MCNP减小计算方差,可通过蒙特卡罗方法的共轭计算来得到共轭通量,经过CADIS方法推导出的公式计算得到权窗参数和源偏倚参数。     

在蒙特卡罗程序中统计共轭通量并计算中子动力学参数的方法

文章介绍了在蒙特卡罗程序中,使用反复裂变几率的统计结果作为共轭通量的估计,并作为权重函数计算动力学参数β_eff和Λ的方法,阐释了在连续能量蒙特卡罗程序MCNP和多群蒙特卡罗程序MCMG中实现这种方法的过程。数值校验结果表明：在几乎不带来附加计算量的同时,在MCMG中使用该方法统计得到的共轭通量与ANISN的共轭通量计算结果符合较好,在MCNP中使用该方法计算得到的中子动力学参数与基准测量结果符合较好。在蒙特卡罗程序中实现了高效率计算中子动力学参数的功能,为蒙特卡罗程序进一步用于反应堆动态行为的分析奠定了基础。     

MC法对移动式工业γ探伤机周边辐射模拟

为准确划定工业γ探伤机的控制区与监督区,应用MCNP5程序构建了移动式γ射线机计算模型。该模型充分考虑了工业γ探伤机散射对周围剂量的贡献,可有效地应用于工业γ探伤机控制区与监督区的划定与环境影响评价。实验结果表明:MCNP程序计算结果与经验公式结果基本一致,MC模拟方法对移动式工业γ射线机周边辐射剂量估算是可行的。     

Characterization of an Am-Be PGNAA set-up developed for in situ liquid analysis: Application to domestic waste water and industrial liquid effluents analysis

A prompt gamma neutron activation analysis (PGNAA) set-up with an Am-Be source developed for in situ analysis of liquid samples is described. The linearity of its response was tested for chlorine and cadmium dissolved in water. Prompt gamma efficiency of the system has been determined experimentally using prompt gamma of chlorine dissolved in water and detection limits for different elements have been derived for domestic waste water. A methodology to analyze any kind of liquid is then proposed. This methodology consists mainly on using standards with water as bulk or in the case of absolute method, to use gamma efficiency determined with prompt gammas emitted by chlorine dissolved in water. To take into account the thermal neutron flux variations inside the samples, flux monitoring was carried out using a He-3 neutron detector placed at the external sample container surface. Finally, to correct for the differences in gamma attenuation, average gamma attenuations factors were calculated using MCNP5 code. This method was then checked successfully by determining cadmium in industrial phosphoric acid and our result was in good agreement with that obtained with inductively coupled plasma (ICP) method.     

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF₃ detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample’s surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.     

Detection limits of pollutants in water for PGNAA using Am-Be source

A basic PGNAA facility with an Am-Be neutron source is described to analyze the pollutants in water. The properties of neutron flux were determined by MCNP calculations. In order to determine the efficiency curve of a HPGe detector, the prompt-gamma rays from chlorine were used and an exponential curve was fitted. The detection limits for typical water sample are also estimated using the statistical fluctuations of the background level in the areas of recorded the prompt-gamma spectrum.     

体源探测效率计算及修正方法研究

用MCNP程序计算HPGe γ探测器与体源有一定距离时的探测效率,该方法对于体源探测效率的刻度有着重要意义。但是由于模型的不确定因素影响,计算结果可能有一定的偏差,充分考虑了影响偏差的因素,采用比较校正法推导了HPGeγ探测器计算体源探测效率修正数学公式。通过实验比较和理论比较方法验证修正公式,比较结果表明：MCNP程序计算体源探测效率可以作为体源探测效率刻度的有效手段。     

Pulse height distributions from deterministic radiation transport simulations

In this paper, we demonstrate that deterministic transport simulations can be used to calculate the pulse height distribution for photons interacting in detector geometries. Typically, Monte Carlo transport methods are used for this application. Utilizing the collided components of the scalar flux calculated from the standard Source Iteration procedure used in many deterministic transport codes, we have generated photon pulse height distributions that compare favorably with those from MCNP5. Several test problems in 1-D slab geometry form the basis of our comparison with MCNP5, but the algorithm is easily extensible to 3-D geometries.     

SPRR-300反应堆混凝土屏蔽层内中子注量率分布研究

采用MCNP程序与ANISN程序结合的计算方案获取了SPRR-300反应堆混凝土屏蔽层内的中子注量率分布情况，同时采用固体核径迹探测器测量了混凝土屏蔽层外低水平中子注量率，两者吻合较好，说明了计算结果的可信性。上述结果为反应堆退役工作提供了放射性源项的计算依据。     

核活化法反应计算中的蒙特卡罗模拟

用MCNP/4B程序,对核活化法反应计算中的截断能量、通量计数方式、数据库、阈探测器之间的扰动等因素的影响作了分析.截断能量可选为有效反应阈能,不影响计算结果,但可减少计算时间.栅元通量计数方式稳定可靠,效率高,计算值略高.不同的数据库的计算结果可能有些差别.阈探测器之间的扰动对计算结果的影响很小.     

Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.     

Application of the Monte Carlo method to the analysis of measurement geometries for the calibration of a HP Ge detector in an environmental radioactivity laboratory

A gamma spectrometer including an HP Ge detector is commonly used for environmental radioactivity measurements. The efficiency of the detector should be calibrated for each geometry considered. Simulation of the calibration procedure with a validated computer program is an important auxiliary tool for environmental radioactivity laboratories. The MCNP code based on the Monte Carlo method has been applied to simulate the detection process in order to obtain spectrum peaks and determine the efficiency curve for each modelled geometry. The source used for measurements was a calibration mixed radionuclide gamma reference solution, covering a wide energy range (50-2000 keV). Two measurement geometries - Marinelli beaker and Petri boxes - as well as different materials - water, charcoal, sand - containing the source have been considered. Results obtained from the Monte Carlo model have been compared with experimental measurements in the laboratory in order to validate the model.     

Application of global variance reduction method to calculate a high-resolution fast neutron flux distribution for TMSR-SF1

The solid fuel thorium molten salt reactor(TMSR-SF1) is a 10-MWth fluoride-cooled pebble bed reactor. As a new reactor concept, one of the major limiting factors to reactor lifetime is radiation-induced material damage. The fast neutron flux(E 0.1 MeV) can be used to assess possible radiation damage. Hence, a method for calculating high-resolution fast neutron flux distribution of the full-scale TMSR-SF1 reactor is required. In this study,a two-step subsection approach based on MCNP5 involving a global variance reduction method, referred to as forward-weighted consistent adjoint-driven importance sampling, was implemented to provide fast neutron flux distribution throughout the TMSR-SF1 facility. In addition,instead of using the general source specification cards, the user-provided SOURCE subroutine in MCNP5 source code was employed to implement a source biasing technique specialized for TMSR-SF1. In contrast to the one-step analog approach, the two-step subsection approach eliminates zero-scored mesh tally cells and obtains tally results with extremely uniform and low relative uncertainties.Furthermore, the maximum fast neutron fluxes of the main components in TMSR-SF1 are provided, which can be used for radiation damage assessment of the structural materials.