镉吸收层球形中子剂量仪能响研究

一、引言 现今国内外使用的慢化型中子剂量当量率仪,其探头结构均为球形或圆柱形,两层慢化体之间的吸收体均用打孔的硼塑料或金属镉。目前最具有代表性的中子剂量当量率仪是瑞典生产的Studsvik 2202D型及英国生产的MK7型。     

用于检测中子辐射的剂量仪

一种用于检测在0．025eV到几百GeV的能量区内的中子辐射的剂量仪。该剂量仪具有用作减速体的基本上呈球形的、包括含氢材料的基体，设置在该基体中心的检测元件，包围该检测元件的中子转换器，该中子转换器包括将所检测的高能中子的能量基本上转换为处于合适能量区域中的中子的金属原子。该剂量仪的特征在于，所述基体具有一个通道，通过该通道可以将检测元件设置到中子转换器中以及从该中子转换器中取出，所述中子转换器构造为圆柱形。     

双球型便携式中子剂量仪探头的设计

介绍了双球型便携式中子剂量仪探头设计的计算过程,利用MCNP3B程序对双球的中子能量响应进行模拟,基于双球的中子能量响应差异,可确定一个修正因子,对H*(10)的估算值进行修正,使得中子剂量能量响应得到改善。考虑到双球的相互影响问题,对双球的间距和角响应进行了计算,并对影响的原因进行了分析。     

可携式中子剂量当量率仪

本文介绍NH1型可携式中子剂量当量率仪的结构和性能。该剂量仪由包有聚乙烯慢化体和硼塑料层的球形~3He计数管和电子学系统组成,适于热能—15MeV能区的中子剂量当量率的测量。剂量当量率测量范围在0.1—10~4μSv/h之间;在热能附近能区和0.3—5MeV能区,剂量当量率响应随能量的变化在±30％以内,对于能量为1.15MeV的中子,响应值约为0.56cps/μSv·h~(-1)。该剂量仪具有模拟和数字显示,重量为7kg。在改换量的名称并对能响作适当修正的条件下,NH1型中子剂量当量率仪可用于中子周围剂量当量率的测量。     

中子剂量当量仪在复杂中子场中的应用研究

针对目前反应堆安全壳内中子剂量测量存在较大误差的特点,通过分析中子剂量当量仪的工作原理、校准方法和现场测量过程中存在的不足,阐述其对测量的影响,并提出修正和弥补措施,以供相关人员参考.     

BH3105E型中子剂量当量仪研制

研制成功BH3105E型中子剂量当量仪.仪器采用镉吸收棒法实现热中子至14 MeV区间等剂量当量测量;选用6Li玻璃闪烁体作为中子探测器;探头输出信号送至硬件电路进行处理.信号经主放大器、脉冲幅度甄别器、整形器由处理器进行计数采集.仪器系统软件设计实现了测量数据存储,采用串口协议与单片机进行通讯.     

一种中子剂量当量仪的探头设计

针对单探测器中子剂量当量仪的能量响应性能普遍不好的情况,设计了一种由三个慢化球壳和三个3He正比计数管构成的多探测器中子剂量当量仪的探头,对它的能量响应进行了计算,并与ICRP的推荐值进行了比较和分析,提出了对它的能量响应性能进行进一步改善的方法。     

一种新型中子剂量当量仪的能量响应性能分析

介绍了对一种新型中子剂量当量仪的能量响应性能的分析.用MCNP程序分别计算了它的各探测单元的中子注量能量响应,并按照中子剂量当量指示值的算法计算了它的中子周围剂量当量能量响应性能,从计算结果看,在热中子～15 MeV的能量范围内它的能量响应变化范围大约在0.55～1.95之间.计算了它在Am - Be源和Cf - 25...     

直读式双量程真空室剂量仪

本文介绍了直读式双量程真空室剂量仪的历史背景、基本结构、工作原理和性能特点。对剂量率效应这一关键性能做了简单的计算和验证。     

高灵敏度环境中子剂量当量仪的优化设计与实验验证

本工作采用MCNP程序对探测器探头的主要参数进行了模拟计算和优化选择。提出了高灵敏度环境中子剂量当量仪的设计方案。最后通过实验验证，试验结果与计算结果基本吻合。     

脉冲中子测量中用长计数器校准闪烁探测器的方法

介绍了采用长计数器与闪烁探测器相组合,通过长计数器对闪烁探测器进行适时校准,进而以闪烁探测器给出低产额脉冲中子注量(或产额)的方法。该方法既减小了长计数器单次计数少引起的统计涨落,又避免了闪烁探测器受γ本底影响大的缺陷,同时还充分利用了长计数器平能量响应以及闪烁探测器探测效率的优点,将长计数器成功地运用于脉冲中子辐射场中,在脉冲中子辐射场中通过长计数器实现了中子注量(率)的量值传递,实现了脉冲中子源中子产额的测量。     

一台测量快中子辐射俘获截面的大液体闪烁探测器

为了测量快中子辐射俘获截面,我们研制了一台测量γ射线的球形液体闪烁探测器,其直径为1m,容积680l。该探测器属4π几何类型,探测器效率达80％以上。为了降低本底,除了用10cm厚的铅和40cm厚的石蜡建成屏蔽室外,还采用符合方法,使本底计数小于100cps。探测器已经在2.5MeV脉冲质子静电加速器上测量了金、钽和铥等元素的辐射俘获截面。     

中子符合计数器在设计与应用时应注意的几个问题

在核材料衡算中,对铀、钚的分析可以采用化学的方法,但需要对样品进行破坏,且操作复杂,测量周期较长。由于铀、钚都可以自发裂变产生中子,可以进行中子测量,中子测量不需要破坏样品,可以对大体积样品进行快速分析,且误差较小。就中子符合计数器进行了参数优化实验,更好地理解了中子探测器的设计原理,包括中子计数统计、屏蔽厚度、不同材料对中子的散射、高压影响、门宽设置、死时间的修正、衰减时间确定、探测效率验证和γ干扰影响等;同时进行了测量条件实验,包括偶然事件影响等。为更好地使用此技术,具有一定的借鉴意义。     

Development of a spherical tissue equivalent proportional counter for neutron monitoring

A spherical tissue equivalent proportional counter(TEPC) for neutron monitoring has been developed. It was properly designed to produce a uniform electric field intensity around the anode wire. An internal ~(241)Am alpha source was adopted for lineal energy calibration. The TEPC was characterized in terms of dose equivalent response in a standard ~(252)Cf neutron field, and was tested with 2.45 MeV neutrons. Microdosimetric spectra, frequency mean lineal energy and dose-average mean lineal energy of 2.45 MeV neutrons were obtained and compared with FLUKA Monte Carlo simulation results. The measurement and simulation results agreed well. The mean quality factor and dose equivalent values evaluated from the 2.45 MeV neutron measurement were in good agreement with the recommended effective quality factor and ambient dose equivalent H*(10),respectively. Preliminary results have proved the availability of the developed TEPC for neutron monitoring.     

单球多计数器的中子能量响应计算

根据球体内随深度变化时中子的慢化程度有所差异．按两两垂直的方法把三个位置灵敏正比计数器安装在一个慢化球体内。用MCNP4A程序计算了6种慢化球体和6种照射方向的能量响应,同时对球半径方向两种分区方法的计算结果进行分析和比较。     

二元慢化型高能中子剂量探测的影响因素研究

为测量高能辐射场及宇宙射线中的中子剂量,基于SMMC设计,对二元慢化型中子剂量探测器进行改进。在慢化球体中加铅等高原子序数材料层,高能中子与铅产生（n,xn）核反应,可提高高能中子的探测灵敏度,使分析的剂量单调能量上限可达10GeV。利用MCNP和FLUKA程序,对铅层位置、铅层厚度、外计数器位置、慢化体直径等因素进行注量响应计算,最终确定探头设计。     

G-M计数管对14MeV快中子的相对灵敏度

用铅吸收方法确定了几种G-M计数管(Philps)18509计数管、(Philps)18550计数管、J417计数管和FWT GM-2计数管)对D-T 14MeV快中子的相对灵敏度(K_α)。对实验数据的处理采用了与现有方法有所不同的方法,并与现有的方法进行了比较分析。对实验方法和数据处理中的有关问题进行了讨论。     

Development of lightweight portable neutron survey meter

We have developed a lightweight portable neutron survey meter comprising a proportional gas counter containing a mixed gas of methane gas and nitrogen gas for measuring the ambient neutron dose equivalent H^*(10) up to about 20 MeV neutrons intended for use in nuclear power plants, and accelerator facilities. Since no heavy polyethylene moderator is used, the survey meter is only about 2.2 kg in weight causing a weight reduction of 70% or more compared to that of the conventional moderated-type survey meter. The spectrum-weight function, G(E) is adopted for dose conversion.

In order to evaluate the energy characteristics of the developed survey meter, mono-energetic neutron reference fields and continuous energy neutron reference fields were used. The evaluation was also carried out by the PHITS calculation. Although the neutron energy response to the mono-energetic neutron fields was highly deviated from the ambient dose equivalent H^*(10) in keV energy region, the response to H^*(10) showed very good agreement only within 25% to the continuous energy neutron fields which are close to the actual work-place neutron fields. The neutron energy response was also investigated for high energy quasi-mono-energetic neutron fields and showed much better results compared with those of conventional moderated-type neutron survey meters.     

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.