Monte Carlo simulation of the NaI(Tl) detector response to measure gold activated foils

This work deals with the implementation of a NaI(Tl) detector for the assessment of the specific saturation activities of pure gold foils after neutron irradiation. These gold foils can be placed in the centre of a set of polyethylene spheres with different diameters. This configuration, known as a passive Bonner sphere system, is suitable to measure neutron spectra normally extended over a wide energy range containing up to 11 decades (from thermal to a few MeV), at places where the neutron field is very intense, high frequency pulsed or where it is mixed with an important high-energy photon component. The MCNPX code was used to evaluate the NaI(Tl) responses to different incident photon energies in terms of pulse-height distributions. An experimental validation of the calculated NaI(Tl) responses, using certified standard sources at a given measurement arrangement, indicates that MCNPX is a valid tool for routine calibration and benchmarking studies of this detector. A good agreement is found between the measured pulse-height distributions of the certified standard sources and those obtained from MCNPX simulations. As a preliminary application, a bare disc Au foil was directly exposed to a Bremsstrahlung photon beam at the isocentre of an 18 MV medical LINAC, in order to test the suitability of this activation material to measure the photo-neutrons generated in such facility. Two differentiated main photo-peaks, arising from ¹⁹⁶Au and ¹⁹⁸Au predominant γ-ray emissions, were observed. The two isotopes are produced mainly by the photonuclear, ¹⁹⁷Au(γ, n)¹⁹⁶Au, and radiative capture, ¹⁹⁷Au(n, γ)¹⁹⁸Au, reactions of, respectively, high-energy photons and thermal neutrons on the gold foil. From the measured ¹⁹⁸Au saturation activity, a rough estimation of (378 ± 68) × 10⁴ cm⁻² Gy⁻¹ was derived for the thermal neutron flux within the LINAC treatment room. This value, although being very approximate, is comparable to those reported by other authors for similar LINAC facilities but with different treatment room configurations, nominal acceleration potentials and Bremsstrahlung photon irradiation areas.     

50MeV／u^12C离子实验靶区中子注量率的测量

用阈探测器中子活化法测量了50MeV/u~(12)C离子实验靶区出射的热中子以及E_n分别大于6、11、20、50MeV的中子注量率。     

快中子照相样品散射本底及图像特性的Monte Carlo模拟

模拟了14MeV中子在穿透样品后与闪烁体光纤的作用。对每根光纤中的能量沉积进行了计算,并转换成可见光(496nm)光子数在模拟实验中,分析了影响图像质量的因素。首先计算了散射中子本底与闪烁体和样品(聚乙烯)间距的关系。当间距为厘米量级时散射中子本底对图像的影响很小。其次,计算表明系统对样品的甄别厚度与入射中子总数N有关,在一定范围内近似与logN成线性关系。最后,通过模拟结果给出了理想平行中子束入射情况下系统的平面分辨率。     

薄膜和纤维塑料闪烁体中子能响特性的Monte Carlo数值计算

根据中子与塑料闪烁体的作用机制及塑料闪烁体的荧光转换特性,研究了塑料闪烁体薄膜和塑料闪烁体纤维中子能响特性的Monte Carlo数值计算方法,给出了不同厚度的塑料闪烁体薄膜和不同直径的塑料闪烁体纤维对裂变能区中子的灵敏度计算结果,并对计算结果及其在电流型坪响应裂变中子探测器方面可能的应用进行了初步探讨。     

~(115)In中子非弹性散射截面的实验测量及蒙特卡罗修正

~(115)In是一种重要的活化材料,准确测量它的中子非弹性散射截面数据对中子注量监测具有重要意义。在四川大学原子核科学技术研究所2.5 MV静电质子加速器上,利用核反应D(d,n)~3He产生的单能中子,以~(197)Au作为标准,采用活化法测量了2.95 Me V、3.94 Me V、5.24 Me V能点的~(115)In中子非弹性散射截面。用Monte Carlo程序MCNPX(Monte Carlo N-Particle eXtended)对靶头材料、冷却水层和样品的包层材料等引起的多次散射效应及注量率衰减效应等进行了修正计算,得到最终结果与Loevestam的计算值符合较好,并且实验中可通过减小靶管、靶底衬、水层及样品的包层材料等厚度来减小多次散射效应和自屏蔽效应的影响。     

蒙特卡罗方法在反应堆物理计算中的应用

利用蒙特卡罗 (MonteCarlo简称MC)方法对高通量工程试验堆的堆芯物理进行了计算 ,计算了该堆的 5个临界装置的有效增殖系数keff以及一个实际运行的复杂堆芯中考验回路内的考验燃料元件的中子通量 ,计算结果与实验值符合得很好     

Monte Carlo method for evaluation of surface emission rate measurement uncertainty

The aim of this study is to evaluate the uncertainty of 2πα and 2πβsurface emission rates using the windowless multiwire proportional counter method.This study used the Monte Carlo method(MCM)to validate the conventional Guide to the Expression of Uncertainty in Measurement(GUM)method.A dead time measurement model for the two-source method was established based on the characteristics of a single-channel measurement system,and the voltage threshold correction factor measurement function was indirectly obtained by fitting the threshold correction curve.The uncertainty in the surface emission rate was calculated using the GUM method and the law of propagation of uncertainty.The MCM provided clear definitions for each input quantity and its uncertainty distribution,and the simulation training was realized with a complete and complex mathematical model.The results of the surface emission rate uncertainty evaluation for four radioactive plane sources using both methods showed the uncertainty's consistency En＜0.070 for the comparison of each source,and the uncertainty results of the GUM were all lower than those of the MCM.However,the MCM has a more objective evaluation process and can serve as a validation tool for GUM results.     

反冲质子探测系统中子灵敏度的Monte Carlo算法

在强流脉冲中子测量工作中,常利用PIN探测器测量中子与聚乙烯中的H核发生弹性散射产生的反冲质子.本文利用Monte Carlo技术进行快速计算的反冲质子探测系统中子灵敏度表达式,给出了算法和计算流程.     

计算高可靠性系统失效概率的统计估计蒙特卡罗方法

在相似仿真方法的基础上 ,设计了计算系统失效概率的统计估计蒙特卡罗方法 ,包括直接统计估计和加权统计估计蒙特卡罗方法。介绍了统计估计蒙特卡罗可靠性仿真的基本原理 ,给出了统计估计蒙特卡罗计算方法的无偏估计量和具体算法。同时采用直接仿真方法、限制抽样蒙特卡罗方法、强迫转换蒙特卡罗方法、直接统计估计和加权统计估计蒙特卡罗方法计算了一高可靠性系统的失效概率 ,结果表明 ,在高可靠性系统不可靠度计算中加权统计估计蒙特卡罗方法计算结果的方差最小 ,效率最高。     

BF3中子探测器阵列探测效率的蒙特卡罗计算

在用241Am-Be中子源对BF3中子探测器阵列探测效率标定的基础上,用蒙特卡罗方法对其探测效率进行了模拟计算,获得了比较满意的结果.然后用蒙特卡罗方法对BF3中子探测器阵列的探测效率进行了研究.研究结果表明,焦面探测器具有较好的探测效率.     

MC模拟能谱对G函数法测量剂量率值结果的影响

为准确、有效地解决固定式剂量仪器的量值溯源,构建基于G函数法的环境级别的剂量装置模型。通过使用MC法,对一款Na I(Tl)探测器的脉冲幅度分布谱进行模拟,并据此考察了截断能量Emin、阶数K、道宽ΔE三个因素对G函数及其计算结果的影响,发现截断能量Emin、阶数K、道宽ΔE对G函数的形状及剂量率计算结果均会产生影响。     

Analytical and Monte Carlo computations of the energy losses by ionisation and their fluctuations in He

Computer programs have been developed to compute the energy loss by ionisation, its fluctuations, spatial distribution of the primary ionisation and cluster distribution, based on analytical and Monte Carlo methods. Results are presented for ⁴He in normal conditions of pressure and temperature and for ³He at P=4 atm and t=25C at intermediate energies, about 150 MeV, and minimum ionisation for pions, protons and deuterons.     

计算维修不独立马尔可夫系统瞬态不可用度的六种蒙特卡罗方法

在马尔可夫可维修系统状态转移积分模型的基础上 ,给出了维修不独立马尔可夫系统瞬态不可用度的六种估计量 ,结合状态转移时间偏倚抽样技术 ,给出了计算维修不独立马尔可夫系统瞬态不可用度的六种蒙特卡罗方法。用两个实际算例考察了各种计算方法的效率随系统运行时间的变化。给出了各种算法适用范围的结论。     

中国先进研究堆标准燃料组件三维流场分析

采用计算流体力学软件CFX4.4和CFX5.5对中国先进研究堆标准燃料组件流场进行了数值模拟。计算得到了额定工况下标准燃料组件内各个冷却剂通道的流量分布和不等间隙通道燃料板两侧压差。根据不同流量下的压降计算结果,给出了标准燃料组件的阻力特性曲线,并与试验结果进行了比较,符合较好。     

Monte Carlo Analysis of Streaming Effect through Annular Gaps around Piping Legs in Sodium Pool to Secondary Sodium Activation of the Demonstration Fast Breeder Reactor

A top-entry loop-type reactor is one of the favorable options for Demonstration Fast Breeder Reactor (DFBR) which is now under development in Japan as a part of conceptual design study. Annular gaps around top-entry piping raise neutron streaming in upward direction in the sodium pool of reactor vessel. It enhances neutron flux level around decay heat exchangers (DHX's) in the pool and at the penetrations in primary biological shield for piping connecting to the vessels of intermediate heat exchangers (IHX's), and consequently enhance secondary sodium activations in these heat exchangers, which is one of the main issues for shielding design. In this study, three-dimensional Monte Carlo analysis method was applied to make precise evaluation of the neutron streaming effect with combination of some techniques for reduction of statistical error within reasonable CPU time. It is established that the contribution of neutron streaming to the secondary sodium activation in DHX's hardly reaches a level of 30% and that design analysis with two-dimensional discrete ordinates method gives conservative evaluation of the secondary sodium activations in DHX's and IHX's.     

Combining Monte Carlo simulations and dosimetry measurements for process control in the Tunisian Cobalt-60 irradiator after three half lives of the source

To check the dose uniformity and to determine the efficiency of medical devices sterilization by gamma irradiation after three half lives of the source,calculations of the absorbed dose were carried out.Monte Carlo simulations and dosimetry measurements,were established to study the radiation processing quality control.An isodose chart was created by GEANT4 Monte Carlo code to evaluate the absorbed dose rate uniformity inside the irradiation room from the year of the installation until the year of the source reload.The dose uniformity ratio (DUR) is deduced from maximum and minimum experimental doses in medical devices after three half lives of the source.     

Development of a practical Monte Carlo based fuel management system for the Penn State University Breazeale Research Reactor (PSBR)

A practical fuel management system for the he Pennsylvania State University Breazeale Research Reactor (PSBR) based on the advanced Monte Carlo methodology was developed from the existing fuel management tool in this research. Several modeling improvements were implemented to the old system. The improved fuel management system can now utilize the burnup dependent cross section libraries generated specifically for PSBR fuel and it is also able to update the cross sections of these libraries by the Monte Carlo calculation automatically. Considerations were given to balance the computation time and the accuracy of the cross section update. Thus, certain types of a limited number of isotopes, which are considered “important”, are calculated and updated by the scheme. Moreover, the depletion algorithm of the existing fuel management tool was replaced from the predictor only to the predictor-corrector depletion scheme to account for burnup spectrum changes during the burnup step more accurately. An intermediate verification of the fuel management system was performed to assess the correctness of the newly implemented schemes against HELIOS. It was found that the agreement of both codes is good when the same energy released per fission (Q values) is used. Furthermore, to be able to model the reactor at various temperatures, the fuel management tool is able to utilize automatically the continuous cross sections generated at different temperatures. Other additional useful capabilities were also added to the fuel management tool to make it easy to use and be practical. As part of the development, a hybrid nodal diffusion/Monte Carlo calculation was devised to speed up the Monte Carlo calculation by providing more converged initial source distribution for the Monte Carlo calculation from the nodal diffusion calculation. Finally, the fuel management system was validated against the measured data using several actual PSBR core loadings. The agreement of the predicted core excess reactivities and the measured values is found to be good considering the measurement uncertainties.     

海底原位PGNAA探测含水率对中子分布影响研究

运用MC法模拟14 MeV快中子进入沉积物后的物理过程,得到了不同深度下沉积物中子能谱分布,分析了0~1 eV中子在沉积物中横、纵向分布规律以及含水率与Cl~-浓度对中子分布的影响,探讨了模拟条件下0~1 eV中子扩散的最大深度范围。结果表明:沉积物含水率对中子能谱分布和0~1 eV中子横、纵向分布均有显著影响,进行中子活化的最佳深度为2 cm处;0~1 eV中子扩散的最大深度范围为20~40 cm;海水中Cl~-浓度对中子吸收有影响但影响不显著。     

PWR（U）乏燃料中超铀元素在混合堆快裂变包层内嬗变研究

从中子学角度对ＰＷＲ（Ｕ）乏燃料中的超铀元素（２３８Ｐｕ，２３９Ｐｕ，２４１Ｐｕ，２４１Ａｍ，２４３Ａｍ，２３７Ｎｐ，２４４Ｃｍ）在聚变－裂变混合堆快裂变包层内嬗变的可行性进了研究。利用一维中子输运和燃耗计算程序ＢＩＤＥＣＡＹ译不同燃料组分的四个快裂变包层进行分析计算。结果表明，在聚变－裂变混合堆快裂变包层内安全，高效地嬗变ＰＷＲ（Ｕ）乏燃料中的超铀元素是可能的。