加速器驱动的次临界系统散裂靶泄漏中子谱研究

为研究加速器驱动的次临界系统（ADS）散裂靶的散裂中子学特性,采用Geant4计算不同能量质子轰击铅铋靶产生的泄漏中子产额、能谱、轴向积分分布。模拟得到1 GeV质子对应的靶的优化尺寸及优化后泄漏中子谱,计算结果可为ADS散裂靶件和堆芯设计提供参考。     

用DCE/CEM和LAHET软件计算ADS标准散裂靶的比较

利用蒙卡程序DCM／CEM和LAHET对加速器驱动的次临界系统ADS标准散裂中子靶进行了计算。长为0．6m、直径为0．2m的圆柱形^208Pb靶在0．15—1．6GeV的高能质子轰击的情况下,利用两软件对质子在靶内的能量沉积、Pb靶发生散裂反应产生的中子及靶内和表面的中子注量、中子能谱分布等进行了比较计算。结果表明：两者模拟结果在中能区(0．8—1．0GeV)符合很好,但在质子能量较高或较低时,两者略有差别。     

铅散裂中子靶物理特性研究

利用蒙卡程序DCM/CEM对ADS标准散裂中子靶进行了计算。计算了长0.6 m,直径0.2 m的圆柱形208Pb靶,在0.1～1.6 GeV的质子轰击下,Pb靶发生散裂反应产生的中子产额及表面的中子注量、能谱分布以及靶内能量沉积分布,解释了以前的实验结果。计算结果与文献数据、实验数据进行了比较,符合良好,对进一步进行ADS堆芯设计具有较好的理论指导意义。     

用不同程序模拟计算ADS散裂靶结果的比对(英文)

利用不同的程序对ADS散裂靶进行了模拟计算。用SNSP,SHIELD,DCM\CEM(Dubna Cascade Model\Cascade EvaporationMode)和LAHET等程序计算了长60 cm,直径20 cm的圆柱形铅靶,分别在800,1000,1500 and 2000 MeV的质子轰击下所产生的泄漏中子产额和能谱分布。模拟结果与实验数据进行了比较,对泄漏中子产额而言,SNSP模拟的结果与实验符合较好,SHIELD,DCM\CEM和LAHET计算的泄漏中子能谱分布比较一致。     

散裂中子源(3.7GeV p+Pb)靶散裂研究中Lavsan探测器的应用

利用Lavsan探测器探测了直径8cm，厚20cm的铅靶中散裂反应的分布规律，并利用理论对实验结果进行了模拟解析，同时利用实测的散裂产物在靶区的分布验证了Lavsan探测结果及理论模拟结果。     

用LAHET和MCNP程序研究散裂中子靶的中子学性能

利用LAHET和MCNP程序对ADS散裂中子靶进行模拟计算。因靶的基本物理性质随束流和靶形状的变化而改变,所以首先评估了源强和靶的几何形状对靶性质的可能影响,然后计算长1.2m、直径为0.6m的圆柱形液态铅靶在1GeV质子轰击下,靶内中子的产生和泄漏及能量的沉积等。与文献数据、实验数据进行了比较,符合良好。计算结果还表明:源强和几何的选择对中子产生和泄漏可产生较大影响;用液态铅作散裂靶时,中子产额和泄漏额较高,且泄漏能谱在可利用范围内,但能量沉积在靶中的分布极不均匀,这可能给传热带来问题。     

0.65～1.5 GeV质子轰击厚Pb靶的中子学研究

在俄罗斯杜布纳联合核子研究所新建高能重离子加速器NUCLOTRON上用0.65～1.0 GeV质子轰击厚Pb靶,对靶区的中子参数进行了研究。使用的比靶的厚度为20cm,直径8cm,呈圆柱状。Pb靶的外围用6cm厚石腊作慢化体。质子沿比靶轴线入射,模拟加速器驱动次临界反应堆（洁净核能系统）的堆芯结构和核反应过程。在Pb靶表面和慢化体表面放置1mm厚CR-39固体径迹探测器薄片作为中子探测器,入射的质子数目为1.55×1013（0.65GeV）、0.84×1013（1.0 GeV）和1.22×1013（1.5 GeV）,用70℃ 6.5mol/L NaOH蚀刻CR-39中的     

快中子能谱在线测量的反冲质子望远镜响应模拟研究

快中子能谱是基于散裂中子源开展大气中子单粒子效应研究的关键输入参数,在线测量宽能区快中子能谱在近散裂靶位置面临飞行时间法不确定度大、中子通量高、本底干扰强等问题。设计了反冲质子望远镜(RPT)系统,利用Geant4模拟了20～200 MeV中子轰击不同厚度聚乙烯转换靶产生的反冲质子产额、角分布以及能谱,为优化探测系统设计提供了指导依据。通过模拟硅探测器与新型快响应CLLB闪烁体组成的二重符合RPT系统对入射中子的响应,分析了影响探测系统探测效率和能量分辨率的因素,确定了聚乙烯转换靶厚度为1 mm、符合质子探测器摆放角度为26.6°和探测器尺寸等重要参数,得到了RPT系统的中子响应函数矩阵,并计算了其探测效率达10-5,对高中子通量和复杂本底干扰环境下的快中子能谱在线测量具有指导意义和参考价值。     

加速器驱动洁净核能系统散裂靶辐射损伤研究Ⅰ.原子位移

考虑了不同入射能量的质子和中子轰击W、Pb靶,利用SHIELD程序系统,研究了我国未来可能使用的靶体的辐射损伤截面、原子位移截面和原子位移率,并同Wechsler的研究结果进行了比较.同时,对1.6 GeV的质子沿中心轴线入射长60 cm、直径20 cm厚铅靶在不同部位引起的辐射损伤进行了研究,得到合理的结果.     

中国散裂中子源强流质子加速器设计、研制及调试运行

中国散裂中子源（CSNS）是基于强流质子加速器的大科学装置,通过高功率质子束流轰击重金属靶产生高通量中子用于开展中子散射研究,CSNS是世界上第四台、发展中国家第一台脉冲型散裂中子源。CSNS包括高功率强流质子加速器、中子靶站和中子谱仪以及相应的配套设施等。加速器由80 MeV负氢直线加速器、1.6 GeV快循环同步加速器及相应的束流输运线组成。CSNS加速器是我国第一台中高能强流高功率质子加速器,本文将介绍CSNS加速器的设计、关键技术、设备研制以及束流调试过程和其中关键问题。     

基于球谱仪的中子剂量率仪设计

研制了一套基于球谱仪的中子剂量率仪。采用单个慢化球及位置灵敏计数器的探测结构,通过单次测量即可计算出中子剂量等信息。在基于电荷分配法的读出电路中,通过对电荷灵敏放大器及滤波成型电路的参数进行优化,使探测器系统的平均位置分辨达到6 mm。利用硬件寻峰等数字信号处理技术,使数据获取系统的最高计数率达到200 k·s-1,满足剂量率仪实时性的要求。     

Recent neutron sputtering results and the status of neutron sputtering

Extensive (d, t) and (d, Be) fast neutron sputtering results obtained by Batelle-Northwest are summarized for Nb and Au. The measured sputtering ratios are in the range of 10^?5 atoms/fast neutron. These results are compared to other available experimental results and to theoretical estimates. Some implications of fast neutron sputtering for future reactors are summarized.     

Study of the neutron multiplication effect in an active neutron method

The neutron multiplication effect appears when an item contains large amounts of nuclear material. The neutron multiplication effect in this paper means the effect of subsequent fission reactions which are caused by fission neutrons produced by interrogation neutrons from a neutron generator. The previous active neutron method could not distinguish between first-fission and subsequent-fission neutrons and might overestimate the amount of nuclear material. However, the neutron multiplication effect in the active neutron method has not been adequately investigated. We discuss the evaluation method of the multiplication effect in the fast neutron direct interrogation method, one of the active neutron methods, using simulations with the Monte Carlo code MVP and experiments involving uranium waste drums. The first-generation neutrons from an external neutron source generate fission neutrons called second-generation neutrons, the second-generation neutrons generate third-generation neutrons, and so on. This study supposes that the neutron multiplication effect is mainly caused by the third-generation neutrons under the condition that the fourth-generation neutrons are much fewer. This paper proposes a correction method for the neutron multiplication effect in the measured data.     

组合PIN脉冲中子探测器的能量响应

组合PIN脉冲中子探测器由一对PIN半导体和中子辐射体组成。理论计算了CH2和^235U分别作为中子辐射体的组合PIN脉冲中子探测器的灵敏度随中子能量的变化;用脉冲中子源实验测量了由CH2辐射体组成的探测器的14MeV中子灵敏度,实验结果与计算符合很好。     

Design of the thermal neutron beam for neutron radiography at the Syrian MNSR

Neutron beam design was studied at the Syrian reactor (MNSR, 30 kW) with a view to generating thermal neutron beam in the vertical irradiation sites for neutron radiography. The design of the neutron collimator was performed using MCNP4C and the ENDF/B-V cross-section library. Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, >10 keV, respectively. To produce a good neutron beam quality, bismuth was used as photon filter. In this design, the L/D ratio of this facility had the value of 125. The thermal neutron flux at the beam exit was about 2.548 × 10⁵ n/cm² s. If such neutron beam were built into the Syrian MNSR many scientific applications would be available using the neutron radiography.     

Evolution of the neutron sensor characteristics in the RBMK-1500 reactor neutron flux

Variation of characteristics of the RBMK-1500 reactor radial neutron flux sensors with the HfO₂ emitter during long-term maintenance was investigated. The influence of nuclear fuel enrichment and burnable erbium admixtures on the energy neutron spectrum, neutron absorption, and hafnium isotopic composition variation was considered. The dependences of corrective factors of the neutron sensor signal on the nuclear fuel burnup depth and the integral current accumulated by the sensor for different enrichment nuclear fuel are presented in the work. The experimental verification of the calculated dependence of the sensor corrective factor on the accumulated integral current was performed.     

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.