^238U裂变电离室修正因子及探测响应的蒙特卡罗模拟计算

利用蒙特卡罗(MC)粒子输运程序MCNPX分别计算了用238U裂变电离室测量由加速器产生的14.8 MeV和25.5 MeV准单能中子注量率以及将其推广应用于测量散裂中子源和宇宙中子源的中子注量率时,由电离室结构、电离室气体、空气等引起的对探测器裂变计数率的修正因子,并给出了探测器在各种情况下的探测响应.为解决蒙特卡罗模拟中探测片太薄、统计误差过大的问题,计算中采用了Dxtran球和强迫碰撞两种方差减少技巧,以降低统计误差、提高计算效率.对于源中子谱覆盖范围较宽、抽样效率低的情况,采用了高能和低能两部分能谱分别计算的方法,以提高计算效率.将模拟计算得到的修正因子应用于探测响应的理论公式,得到相应的探测响应,并与MC模拟计算直接得到的探测响应进行了比较,对模拟计算进行自洽性验证.利用伴随α粒子测量装置和电离室同时测量14.8 MeV准单能中子注量率,得出238U裂变电离室对串列加速器上14.8 MeV准单能中子场的探测响应,与MC模拟计算结果进行比较,对模拟计算进行实验验证.     

液体闪烁探测器中子γ射线测量研究

液体闪烁体对中子的响应函数,利用O5S蒙特卡罗程序进行计算,γ射线响应函数利用MATHA蒙特卡罗程序进行计算。分别对液体闪烁探测器(n,γ)分辨品质测量、Am-Be中子源的中子谱和γ射线能谱的测量、铁球在D-T中子照射下泄漏中子和伴生γ射线能谱的测量、水球在D-T中子照射下泄漏中子和伴生γ射线能谱的测量等内容进行了介绍,并对结果进行了分析和讨论。     

快中子诱发~(238)U裂变中子能谱的核-4乳胶测量

介绍了用核-4乳胶测量2.5 MeV中子诱发~(238) U裂变瞬发中子谱的实验条件和布局以及核乳胶中反冲质子径迹的显微镜读取和处理,并采用蒙特卡罗程序SRIM和拟合方法导出了该类乳胶中反冲质子的能量-射程解析关系式,进而反推得到了核乳胶处的裂变中子相对能谱,并进行了修正,最后将实验结果与经典的Maxwell理论模型计算谱进行了比较。结果表明,二者在不确定度范围内符合较好,验证了核-4乳胶在测量裂变中子谱高能段的有效性。     

液体闪烁探测器中子和γ射线测量研究

液体闪烁体对中子的响应函数,利用05S蒙特卡罗程序进行计算,γ射线响应函数利用MATHA蒙特卡罗程序进行计算。分别对液体闪烁探测器（n,γ）分辨品质测量、Am—Be中子源的中子谱和γ射线能谱的测量、铁球在D—T中子照射下泄漏中子和伴生γ射线能谱的测量、水球在D—T中子照射下泄漏中子和伴生γ射线能谱的测量等内容进行了介绍,并对结果进行了分析和讨论。     

基于VENUS-2临界基准的CosMC程序验证

堆芯中含有大量经过多个循环、燃耗较高的钚燃料时,堆芯中子学特性会发生变化。为了验证目前核数据库及现有程序对这种情况的计算精度,经济合作与发展组织核能机构(OECD/NEA)提出了VENUS-2基准实验。Cos MC程序是专门用来进行反应堆计算的蒙特卡罗程序,可以处理复杂几何模型。本文采用最新核数据库及Cos MC程序对VENUS-2基准进行了计算,计算结果与其他程序做了对比,结果表明:Cos MC的计算结果与实验测量值以及其他程序计算的部分结果符合的较好,说明用Cos MC程序计算含混合氧化物(MOX)燃料堆芯的临界问题是可行的。     

SPRR-300反应堆大热柱内中子注量率及能谱分布

利用MCNP程序校核ANISN程序计算出的堆芯分布，进行一维空间简化的修正；同时采用延伸横向尺寸的方法近似替代无法在一维模型中建立的反射层，以进行横向中子泄漏修正。经此修正后，用一维ANISN程序计算了SPRR-300反应堆热柱内的中子注量率分布和中子能谱。热柱内镉比的程序计算值与实验测量结果基本一致，两者间的偏差在5%以内，个别位置处的偏差不大于10%。这一结果表明，对热柱内中子注量率分布及能谱等深穿透问题，采用确定论一维离散程序ANISN可获得很好的计算结果。     

三维离散纵标方法在堆内构件释热率计算中的初步应用

以美国H.B.Robinson-2#机组反应堆压力容器(RPV)基准实验的参数为输入数据,采用三维离散纵标方法程序(TORT)计算压力辐照监督管处中子能谱及典型核素的活度值。计算得到的辐照监督管处中子能谱与基准实验结果趋势一致、吻合较好;典型核素活度的计算值与测量值之比(C/M)为1.04±0.04。用TORT对福建宁德核电站堆内构件释热率分布进行初步计算,并与蒙特卡罗方法(MCNP)的计算结果相比较,两种方法的结果表现出良好的一致性。最后对TORT程序应用于堆内释热率计算进行讨论。     

钒球14 MeV中子的泄漏能谱测量

建立了厚度为 1 0 5cm的金属钒球基准实验装置。钒的纯度为 99 9%。用NE 2 1 3谱仪测量了d T中子的 0 75～ 1 5MeV泄漏中子能谱 ,能量大于 0 75MeV的中子的穿透率为 0 84± 0 0 3 ,中子能谱实验误差为 5 %～ 7%。用MCNP/ 4AMonte Carlo程序和FENDL 2库核数据进行了模拟计算 ,并与实验结果进行了比较     

锰浴法测量中子源发射率中各修正因子的蒙特卡罗计算

在采用锰浴法对放射性同位素中子源的发射率进行测量时,MnSO₄溶液中⁵⁵Mn仅俘获部分源中子,故需考虑对未被⁵⁵Mn俘获部分的修正。用蒙特卡罗粒子输运程序MCNP对中子源强标准装置的锰浴系统进行模拟计算,对实验测量结果进行修正,并通过由国际计量局（BIPM）组织的中子源强度国际比对,验证了计算所建立模型的可靠性。利用MCNP程序的微扰计算功能,通过考虑MnSO₄溶液的密度、源及其承托物材料的密度、源的位置、锰浴半径以及反应截面的改变对计算结果的影响,对模拟计算结果的不确定度进行了详细评定,提供了一种蒙特卡罗模拟计算结果不确定度的评定方法。     

基于堆外中子能谱测量的堆芯中子场反演方法

目前商用反应堆堆内采用的自给能探测器(Self-Powered Detector,SPD)可以实现堆芯中子通量分布的在线测量,但难以测量中子能谱。同时由于一些先进反应堆堆内服役环境更加恶劣,现有的堆内测量系统难以满足先进反应堆堆芯中子场在线测量的需求。本文提出了一种基于堆外中子能谱测量的堆芯中子场反演方法,通过正向中子输运计算构建堆芯相邻截断面之间的能谱响应矩阵,实现反应堆堆芯中子场的反演计算。通过采用两个简化的反应堆模型进行验证,其中对压力容器处反演的中子能谱与蒙特卡罗输运计算结果平均相对偏差约为14%,在外层燃料组件区域反演的中子能谱与蒙特卡罗输运计算结果平均相对偏差约为11%,初步验证了本方法的可行性。     

基于反应性温度系数的金属型脉冲堆波形计算

金属型脉冲堆的反应性反馈效应主要由热膨胀引起,本文在反应性温度系数的基础上建立了波形计算方法,该方法由蒙特卡罗中子输运程序、热力学计算程序和点堆方程3部分组成。首先由三维中子输运程序和热力学计算程序计算出热功率和反应性的耦合关系,然后将耦合关系代入点堆方程,即可求解出波形。采用该方法计算了Lady Godiva的波形,计算结果与LANL的实验结果一致。     

基于IQS/MC方法的ADS次临界反应堆中子时空动力学模拟分析

采用改进准静态近似与蒙特卡罗中子输运程序相结合(IQS/MC)的方法实现了加速器驱动的次临界系统(ADS)中子时空动力学模拟计算。以加速器驱动嬗变研究装置的靶堆耦合参考方案物理模型为例,通过对束流瞬变引入和燃料组件提升两种工况进行动态模拟,计算得到了堆芯总的相对功率、分能群相对中子注量率及相对功率三维网格分布随时间的变化。将IQS/MC方法计算结果与点堆计算结果进行了对比分析,模拟结果符合物理规律,两种方法对比结果与国外相关文献一致,表明IQS/MC方法适用于ADS次临界反应堆中子时空动力学过程的瞬态安全分析。     

ADS次临界系统中子时空动力学计算与瞬态分析

加速器驱动次临界反应堆(ADS)中子时空动力学计算需要考虑外中子源和空间分布的影响,比临界系统中子动力学计算要复杂得多。本文将改进准静态(IQS)近似与蒙特卡罗(MC)方法相结合,对于带外源的ADS次临界系统中子时空动力学过程,形状函数、动力学参数由MCNPX程序计算得到,幅度函数与集总参数热工反馈模型进行耦合计算,并开发了IQS/MC计算程序可视化操作界面。针对CIADS靶堆耦合系统参考方案物理模型,对引入束流瞬变及无保护失流工况过程进行瞬态模拟计算分析,给出了堆芯相对功率、燃料温度及冷却剂出口温度随时间的变化曲线。同时,将中子注量率进行分群计算,得到了堆芯分能群的相对中子注量率网格分布随时间的变化,模拟结果与理论分析一致。     

90°伴随粒子法D-D中子产额测量修正因子计算

开展了90°伴随粒子法D-D中子产额测量中的修正因子理论计算方法研究,基于MATLAB软件平台,开发了用于修正因子计算的计算机程序.计算给出了厚靶条件下,入射氘能量在20-700 keV范围,90°伴随粒子法D-D中子产额测量各向异性修正因子Rthick、中子和质子产额比(Yd,n/Yd,p)thick及总修正因子RY,并与早先的研究结果进行了对比,分析了计算结果的不确定度,总修正因子计算数据的不确定度约为2％.     

Study on kinetics of fast subcritical assembly—Contribution of delayed neutrons to the transition after a reactivity insertion in a subcritical system—

This study is an investigation of the effect of the delay neutron on the kinetics in the subcritical system. And, it proposes a method necessary for the kinetics code development that uses the Monte Carlo (MC) computation.

It is generally difficult to analyze three dimensional space and time dependent kinetics by using a MC method. It is because the sampling of the neutron in a region becomes difficult when conditions of the region changes with time. In this study, we consider about the effect of delayed neutron in the kinetics of ADS. The behavior of neutrons is considered spontaneous in this system. It means a neutron is absorbed or leaks in a short period, while the conditions of region do not change. Therefore they are treated by steady state calculation. On the other hand the densities of delayed neutron precursors changes slowly, and the conditions of region change. In the concept of developed MC method, the neutrons are calculated by using steady state equation at each time point, and the delayed neutron precursors are calculated by using time dependent equation. We tried to inspect the accuracy of this method by using a point equation. We obtained strict solution Φ* as a reference solution, Φ1 as a solution by the present method, and Φ2 as the solution where both neutrons and delayed neutron precursors are treated by using static equations. The obtained results show a good agreement between Φ1 and Φ*, though the Φ2 agrees with Φ* poorly in all cases. Especially, we showed that this technique was effective from the reactivity change by ADS, and the relation of a delayed neutron. Finally, the effect of the delay neutron on the beam trip in the neutron source for the drive was examined by using the technique of Φ2.     

热中子吸收材料特性模拟计算与分析

以核反应堆为中子源,通过蒙特卡罗(Monte Carlo,MC)方法计算的手段,分析了几种热中子吸收材料的特性,并总结出将热中子吸收材料用于建立超热中子辐射场时的一般规律,从而优选出适用于建立超热中子辐射场的热中子吸收材料.     

Life extension simulation of aged reactor pressure vessel material using probabilistic fracture mechanics analysis on a massively parallel computer

This paper describes a probabilistic fracture mechanics (PFM) computer program using the parallel Monte Carlo (MC) algorithm. In the stratified MC algorithm, a sampling space of probabilistic variables such as fracture toughness value, the depth and aspect ratio of an initial semi-elliptical surface crack is divided into a number of small cells. Fatigue crack growth simulations and failure judgements of those samples are performed cell by cell in parallel. The developed PFM program is implemented on a massively parallel computer composed of 512 processors. As an example, some life extension simulations of aged reactor pressure vessel material are performed, taking analysis conditions of normal and upset operations of PWRs. The results show that cumulative breakage probabilities of the analyzed model are of an order of 10⁻⁷ (1/crack), and that parallel performance always exceeds 90% owing to an employed function of dynamic workload balancing. It is also demonstrated that the degradation of fracture toughness values due to neutron irradiation and the probabilistic variation of fracture toughness values significantly influence failure probabilities.     

最优化方法应用于裂变核反应理论计算的研究

中子诱发裂变核反应的反应机制十分复杂,需由光学模型、复合核模型等共同描述,因此需同时对大量参数进行调节才能正确描述裂变核反应数据。本文在FUNF裂变核反应模型基础上,利用先进的最优化方法MINUIT,结合MPI的并行计算方法,研究了MINUIT对裂变核反应调参的工作原理,并以中子诱发238U的裂变核反应为例,验证了裂变核反应数据优化方法的可靠性。     

Application of linear-extended neutron diffusion equation in a semi-infinite homogeneous medium

The linear-extended neutron diffusion equation (LENDE) is the volume-averaged neutron diffusion equation (VANDE) which includes two correction terms: the first correction is related with the absorption process of the neutron and the second is a contribution to the neutron diffusion, both parameters are related to neutron effects on the interface of a heterogeneous configuration. In this work an analysis of a plane source in a semi-infinite homogeneous medium was considered to study the effects of the correction terms and the results obtained with the linear-extended neutron diffusion equation were compared against a semi-analytical benchmark for the same case. The comparison of the results demonstrate the excellent approach between the linear-extended diffusion theory and the selected benchmark, which means that the correction terms of the VANDE are physically acceptable.     

Averaging the neutron diffusion equation

This paper presents a general theoretical analysis of the neutron motion problem in a nuclear reactor, where large variations on neutron cross-sections normally preclude the use of the classical neutron diffusion equation. A volume-averaged neutron diffusion equation (VANDE) is derived which includes correction terms to diffusion and nuclear reaction effects. A method is presented to determine closure-relationships for the VANDE (e.g., effective diffusivity). In order to describe the distribution of neutrons in a highly heterogeneous configuration, it was necessary to extend the classical neutron diffusion equation. Thus, the averaged diffusion equation includes two correction factors: the first correction is related with the absorption process of the neutron and the second correction is a contribution to the neutron diffusion, both parameters are related to neutron effects on the interface of a heterogeneous configuration. As an example of the VANDE, the plane source in an infinite medium was considered to study the effects of the correction factors on the neutron flux, and the results were compared with classic solution.