Measurement of neutron spectra through composed material block bombarded with D-T neutrons

A 2-dimensional composed material assembly made of the iron and hydric block has been established. The neutron spectra from the assembly bombarded with 14-MeV neutrons at neutron generator have been obtained using the proton recoil technique with a stillbene detector. The detector positions were selected at the 60°, 120°, 180° on the surface of the iron spherical shell. The background neutron spectra consisted of background and room return radiation were subtracted with combination of methods of experimental shielding and MCNP calculation. The uncertainty of results was 6.3-7.4%. The experiment results were analyzed and simulated by MCNP code and two data library. The difference is integral neutron flux (background neutron subtracted) of measured results greater than calculations with maximum of 21.2% in the range of 1-16 MeV.     

可控中子源中子-伽马测井仪中屏蔽结构的优化

本文设计了一种使用遗传算法调用蒙特卡罗计算软件MCNP的方案,用以优化设计中子-伽马测井仪中的屏蔽结构。以D-D聚变中子源和BGO探测器为研究对象,以最小化探测器内的辐照本底为优化目标,设计出了3种不同厚度的屏蔽结构。模拟结果表明,这些屏蔽结构具有优异的屏蔽性能,可有效地降低探测器中的辐射本底。     

The effect of carbon powder addition on the properties of hematite radiation shielding concrete

Carbon powder was added to shielding concrete made of Hematite aggregates to investigate its effects on shielding properties. The powder was added in different percentages, and the mechanical and radiation attenuation properties of the prepared concretes were determined.It was found that, the addition of carbon powder by 6% (by wt.) of the concrete could increase the strength on concrete by about 15%. The shielding effectiveness decreased for both gamma and neutrons with the increasing carbon powder percentage. But the loss in shielding effectiveness for both gamma rays and neutrons were within measurements error range for carbon powder addition of 6%.Simulation for the experimental measurements was carried out using Monte Carlo MCNP code, to understand the effect of carbon powder on the shielding effectiveness against neutrons. The results of the simulation were in good agreement with the experimental results.     

Modeling, analysis and prediction of neutron emission spectra from acoustic cavitation bubble fusion experiments

Self-nucleated and external neutron nucleated acoustic (bubble fusion) cavitation experiments have been modeled and analyzed for neutron spectral characteristics at the detector locations for all separate successful published bubble fusion studies. Our predictive approach was first calibrated and validated against the measured neutron spectrum emitted from a spontaneous fission source (²⁵²Cf), from a Pu–Be source and from an accelerator-based monoenergetic 14.1 MeV neutrons, respectively. Three-dimensional Monte-Carlo neutron transport calculations of 2.45 MeV neutrons from imploding bubbles were conducted, using the well-known MCNP5 transport code, for the published original experimental studies of Taleyarkhan et al. [Taleyarkhan, et al., 2002. Science 295, 1868; Taleyarkhan, et al., 2004. Phys. Rev. E 69, 036109; Taleyarkhan, et al., 2006a. PRL 96, 034301; Taleyarkhan, et al., 2006b. PRL 97, 149404] as also the successful confirmation studies of Xu et al. [Xu, Y., et al., 2005. Nuclear Eng. Des. 235, 1317–1324], Forringer et al. [Forringer, E., et al., 2006a. Transaction on American Nuclear Society Conference, vol. 95, Albuquerque, NM, USA, November 15, 2006, p. 736; Forringer, E., et al., 2006b. Proceedings of the International Conference on Fusion Energy, Albuquerque, NM, USA, November 14, 2006] and Bugg [Bugg, W., 2006. Report on Activities on June 2006 Visit, Report to Purdue University, June 9, 2006]. NE-213 liquid scintillation (LS) detector response was calculated using the SCINFUL code. These were cross-checked using a separate independent approach involving weighting and convoluting MCNP5 predictions with published experimentally measured NE-213 detector neutron response curves for monoenergetic neutrons at various energies. The impact of neutron pulse-pileup during bubble fusion was verified and estimated with pulsed neutron generator based experiments and first-principle calculations. Results of modeling-cum-experimentation were found to be consistent with published experimentally observed neutron spectra for 2.45 MeV neutron emissions during acoustic cavitation (bubble) fusion experimental conditions with and without ice-pack (thermal) shielding. Calculated neutron spectra with the inclusion of ice-pack shielding are consistent with the published spectra from experiments of Taleyarkhan et al. [Taleyarkhan, et al., 2006a. PRL 96, 034301] and Xu et al. [Xu, Y., et al., 2005. Nuclear Eng. Des. 235, 1317–1324] where ice-pack shielding was present, whereas without ice-pack shielding the calculated neutron spectrum is consistent with the experimentally observed neutron spectra of Taleyarkhan et al. [Taleyarkhan, et al., 2002. Science 295, 1868; Taleyarkhan, et al., 2004. Phys. Rev. E 69, 036109] and Forringer et al. [Forringer, E., et al., 2006a. Transaction on American Nuclear Society Conference, vol. 95, Albuquerque, NM, USA, November 15, 2006, p. 736; Forringer, E., et al., 2006b. Proceedings of the International Conference on Fusion Energy, Albuquerque, NM, USA, November 14, 2006] and also that from GEANT computer code [Agostinelli, S., et al., 2003. Nuclear Instrum. Methods Phys. Res. A 506, 250–303] predictions [Naranjo, B., 2006. PRL 97 (October), 149403] in which ice shielding was also absent.The results of this archive confirm for the record that the confusion and controversies caused from past reports [Reich, E., 2006. Nature (March) 060306. news@nature.com; Naranjo, B., 2006. PRL, 97 (October) 149403] have resulted from their neglect of important details of bubble fusion experiments. Results from this paper demonstrate that ice-pack shielding between the detector and the fusion neutron source, gamma photon leakage and neutron pulse-pileup due to picosecond duration neutron pulse emission effects play important roles in affecting the spectra of neutrons from acoustic inertial confinement thermonuclear fusion experiments.     

Study of the spectra and doses created in the iron-water shielding of a monoenergetic neutron source

Many-group calculations were made for the penetration of neutrons, emitted from monoenergetic sources, through water, iron, andwater-iron systems of finite dimensions; the results of these calculations are presented. The neutron spectra resulting from the passage of such neutrons through water and iron shielding layers were calculated on the twenty-group diffusion-transport approximation, Detailed attention was paid to the high-energy part of the spectrum; certain peculiarities in neutron migration and moderation processes in shielding of the type in question were elucidated. Dose curves D(r) were plotted for neutrons of various energies.By using the superposition principle, the results enable the neutron spectrum to be determined for sources having any arbitrary spectrum.Translated from Atomnaya Énergiya, Vol. 21, No. 1, pp. 27–35, July, 1966.     

Analyses of Data Acquired From High-Pressure$hbox Xe+ ^3!hbox He$Gas Ionization Chamber

It is shown that pulse-height spectra acquired from a high-pressure$hbox Xe+ ^3!hbox He$gas ionization chamber exposed to mixed gamma-neutron radiation fields can be unfolded into separate gamma ray and thermal neutron spectra. The procedure takes advantage of the unique shape of the recently discovered spectral response of thermal neutrons in this high pressure$hbox Xe+ ^3!hbox He$mixture. A template spectrum formed from only the pulse-height distribution of neutron signals is subtracted from the combined gamma and neutron spectrum to provide an estimate for the count of thermal neutrons. This procedure leaves a pure gamma spectrum for standard gamma ray spectrum analyses and isotope identification.     

Relation between material configuration and radiation shielding capability - application of radiation shielding characteristic functions

For the purpose of finding a principle for material configuration which an ideal radiation shielding in slab geometry should obey, radiation energy dependence of material configuration is studied. In the course of study, radiation shielding capability for each system of different material configuration is evaluated by using radiation shielding characteristic functions defined as dose rates of transmitted radiations in response to isotropic incidence of radiations to the slab shield with pulse-like narrow energy distributions.In shielding neutrons by steel and water layers, recommendable material configuration depends on energy distribution of incident neutrons; all steel layers should be located in the source side of all water layers, if incident neutron energies are above 5 MeV: either homogeneous array of steel and water layers or above mentioned material configuration is recommendable, if incident neutron energies are between 2 MeV and 5 MeV: all water layers should be located in the source side of all steel layers, if incident neutron energies are below 2 MeV or incident neutrons have energy spectrum of fission neutrons.Above recommendation can be understood well by considering both energy dependence of neutron cross sections of each material and the maximum amount of energy degradation at elastic scattering in each material.In designing a neutron shield, shielding of secondary gamma rays is important as well as neutron shielding. This importance is demonstrated for several types of actual cask walls which are composed of many material layers by using the characteristic functions of neutrons and gamma rays for cask walls.     

基于双GM计数管设计的宽量程伽马剂量计性能初步测试

为在核事故中使用机器人定位和抓取丢失的伽马放射源,本文初步设计了一款基于双GM（Geiger Müller）计数管（ZP1321,ZP1301）的宽量程伽马剂量计。设计了用于探测器信号处理的前端模拟电路和基于微控制器的数据处理和传输程序等,并使用137Cs伽马放射源对设计的系统进行了初步测试。测量并分析了影响剂量计输出信号的主要因素。初步实验结果表明,所设计的无屏蔽壳的伽马剂量计可成功地在线检测伽马放射源。     

Measurement of dose rate profile and spectra through a cylindrical duct vis-à-vis Monte Carlo simulation studies for optimisation of reactor shield design

In the design of a nuclear reactor, penetrations are provided in the top shield to carry out some essential operations. Radiation streaming is envisaged through such penetrations. To avoid radiation streaming, complementary shielding is provided. Optimisation of complementary shielding is carried out by performing calculations using MCNP code. Uncertainties in the calculations are taken care of by incorporating a safety factor. The assumption of the safety factor, while designing the reactor shielding, has been validated by undertaking experimental measurements on a similar geometry vis-à-vis the computed values obtained using MCNP code. The results of the present work agree with the safety factor of two assumed during the shield design. The details of gamma spectral measurements carried out with high purity germanium detector to understand the pattern of the scattered spectrum are also presented.     

贫化铀用于核动力堆防护的屏蔽性能与放射性毒性计算分析

针对贫化铀的γ射线屏蔽进行了实验与模拟计算验证。构建了核动力压水堆屏蔽模型,模拟输出的屏蔽层内中子能谱与实际能谱分布较为一致。采用蒙特卡罗程序与燃耗计算程序相耦合的方法,模拟计算了贫化铀在不同位置处中子、γ混合辐射场中的综合屏蔽性能,并与铅作为屏蔽材料进行了对比分析。模拟计算了屏蔽层中子辐照贫化铀40 a后的活化和裂变产物,分析了材料辐照前后年摄入量限值（ALI）定义下的放射性毒性,结果表明,新增二次产物对放射性毒性影响不大。      

Experiments and Calculations of 14 MeV Neutron Streaming through Multi-Layered Slit Assembly

A streaming experiment using a D-T neutron source was carried out to verify the calculational technique for neutron transport in a shield assembly with multi-layered slits. Reaction rate distributions of a small spherical NE213 scintillation detector to fast neutrons were measured in the slits made of 304SS and in the mortar surrounding the slits. The energy spectrum of fast neutrons in the slit was also measured with the same detector. These measurement were compared with calculations using the continuous energy Monte Carlo transport code MCNP. The calculated reaction rates in the slits agreed with the measured ones within experimental and calculational errors. Besides, it is suggested that the attenuation of fast neutrons in the mortar is significantly different from that in the slits and the behavior is nearly traced by the calculation with the MCNP code. The measured and calculated spectra at a position close to the exit inside the lower slit agreed within the both errors.     

裂变缓发γ射线能谱的蒙特卡罗模拟

采用递次衰变路径搜索和遍历的递归算法编制一程序,该程序可用于计算裂变核素在中子辐照时和辐照后任意1种或1组裂变产物在任意时刻的放射性活度、γ能谱及其随时间的变化。计算了²³⁹Pu在池式堆快中子照射下的裂变缓发γ能谱。用MCNP软件模拟了高纯锗探头对裂变缓发γ射线的能谱响应。模拟结果可用于指导核材料裂变产额测量等研究工作。     

NE213闪烁体的n-γ分辨

NE213闪烁体广泛使用于探测快中子,但却伴随很高的g本底。本文使用快信号门与总信号门的两门积分方法,用快信号与总信号进行二维作图来分辨粒子。用252Cf 中子源和d+D核反应产生的单能中子研究了 f5"2"、f5"5"和f8"2"三种尺寸的NE213闪烁体的n-g分辨性能,通过选择合适的门宽,获得了极好的n-g分辨效果。对f5"2"的NE213,最佳快信号门宽为30ns。 摘要 NE213闪烁体广泛使用于探测快中子,但却伴随很高的g本底。本文使用快信号门与总信号门的两门积分方法,用快信号与总信号进行二维作图来分辨粒子。用252Cf 中子源和d+D核反应产生的单能中子研究了 f5"2"、f5"5"和f8"2"三种尺寸的NE213闪烁体的n-g分辨性能,通过选择合适的门宽,获得了极好的n-g分辨效果。对f5"2"的NE213,最佳快信号门宽为30ns。     

Ge(Li)探测器测定14MeV中子引起~(238)U裂变中稀土核素的裂变产额

在核化学研究中,对稀土核素生成截面的资料是非常感兴趣的。因为在各种能量的轻粒子引起重核的裂变,或者在重离子引起的核反应中,稀土核素都构成了核反应产物中相当大的一部分。此外,处于大的核形变区的稀土核远离任何满核子壳层,因此测量稀土核的反应截面能够获得不受壳效应影响的核反应信息。     

The limitations of associated alpha particle technique for contraband container inspections

Inspection of a shipping container for the presence of the threat materials has been investigated in the laboratory by using a 14 MeV neutron beam, a BaF₂ gamma detector and the associated alpha particle technique. The associated alpha particle technique is proposed as a part of a two sensor system for contraband container inspections. This method is effective in the reduction of background radiation with the possibility of collimating electronically the neutron beam.The intrinsic time resolution has been experimentally estimated to be 1.3 ns (FWHM), which allows inspection of a minimum voxel having 7 cm depth along the neutron flight path. The neutron beam intensity plays a crucial role as a limiting factor for the acquisition time reduction. Single counting rates of the gamma and alpha detector were investigated as a function of the neutron intensity, distance between the gamma detector and the neutron source and the type of shielding. The time and the energy spectra for different neutron intensities were evaluated.     

Components of the background of Ge(Li) and Ge detectors in passive shielding

The components of the background of a Ge(Li) detector and an ultrapure Ge detector in gamma spectrometers in passive shielding with a special design were studied in a ground-surface laboratory in 1996–2006. The measurement time varied from 45 to 240 h. The background for the detectors is due to radionuclides in the shielding material and cavities in the shielding and the detector materials themselves. Special attention is devoted to the study of the time dependence found for the background of the daughter of products of ²²²Rn decay, including the 46.5 keV peak of ²¹⁰Pb. __________ Translated from Atomnaya énergiya, Vol. 103, No. 5, pp. 318–322, November, 2007.     

Comparison of measured and calculated neutron and gamma ray energy spectra behind an in-line shielded duct

Integral experiments that measure the transport of 14 MeV neutrons through a 0.30-m-diameter duct having a length-to-diameter ratio of 2.83 that is partially plugged with a 0.15 m diameter, 0.51 m long shield comprised of alternating layers of stainless steel type 304 and borated polyethylene have been carried out at the Oak Ridge National Laboratory. Measured and calculated neutron and gamma ray energy spectra are compared at several locations relative to the mouth of the duct. The measured spectra were obtained using an NE-213 liquid scintillator detector with pulse shape discrimination methods used to simultaneously resolve neutron and gamma ray events. The calculated spectra were obtained using a computer code network that incorporates two radiation transport methods: discrete ordinates (with P₃ multigroup cross sections) and Monte Carlo (with continuous point cross sections). The two radiation transport methods are required to account for neutrons that singly scatter from the duct to the detectors. The calculated and measured neutron energy spectra above 850 keV agree within 5–50% depending on detector location and neutron energy. The calculated and measured gamma ray energy spectra above 750 keV are also in favorable agreement, 5–50%, depending on detector location and gamma ray energy.     

重屏蔽条件下铀材料丰度的时间关联测量分析

针对高原子序数物质屏蔽下的铀材料,提出了一种新的技术方法来估算其铀丰度。首先,用D-T中子发生器的14 MeV中子对铀材料及屏蔽层进行主动质询,同时利用裂变射线探测器得到裂变中子/γ射线时间关联测量谱。然后,利用成像探测器建立的断层扫描图像得到铀材料及屏蔽层的几何和材料参数,并调用不同的铀丰度参数进行蒙特卡罗模拟计算,得到时间关联计算谱。最后,寻找与测量谱最匹配的模拟谱,确定铀材料的实际铀丰度。通过对比裂变射线探测器实验测量得到的所有裂变中子/γ射线时间关联谱,结果发现在各种屏蔽层状态下,高浓铀材料与贫化铀材料的时间关联谱均存在显著差异,可用时间关联谱作为区分不同铀材料丰度的重要技术特征;对于相同铀丰度和屏蔽状态下的铀材料,模拟谱与测量谱吻合较好,表明时间关联谱的模拟与实验分析可为铀材料铀丰度的估算提供技术基础。     

EJ339A型液体闪烁探测系统光输出响应函数的理论模拟与实验研究

对液体闪烁探测器EJ339A中子与伽马辐射测量问题,采用理论模拟与实验分析方法,结合自主设计搭建的基于Lab VIEW的数字化信号处理系统,分别完成22Na、133Ba、137Cs和60Co等4种不同能量的γ源等效电子能量测量与刻度。在此基础上,利用中子飞行时间测量原理,在不同时间窗下将锎(252Cf)源近似分化成若干个单能中子源,获得1.9 7.8 Me V范围内中子在探测器中的光输出响应函数。结果表明,理论模拟与实验测量值在低能段比较吻合,而高能段存在约7.3%的误差。