中子和伽马射线综合屏蔽材料研究进展

核屏蔽设计的主要任务是屏蔽中子和γ射线,因此,兼具中子和γ射线综合屏蔽性能的屏蔽材料无疑是核屏蔽设计时的最佳选择。本文主要针对中子和γ射线综合屏蔽材料的种类、基质材料和性能进行了分析和综述,探讨了现有中子和γ射线综合屏蔽材料存在的主要问题,认为屏蔽材料的屏蔽性能与其他性能(如力学性能、耐热性等)之间的矛盾是屏蔽材料需要解决的关键问题。最后,展望了未来中子和γ射线综合屏蔽材料的研究方向,指出结构功能一体化屏蔽材料的研究应用是未来一大发展趋势。     

中子屏蔽材料的研究进展

中子屏蔽材料是一种具有快中子慢化性能的基体材料和具有吸收中子性能的吸收材料混合而成的屏蔽中子材料,目前已经广泛应用于医疗、农业、航空、核反应堆等领域。简述了中子屏蔽材料的一般要求,主要论述了多种类型的中子屏蔽材料现状和性能。对新型中子屏蔽材料进行了展望,指出开发具有随意裁剪、任意粘贴与包装性能与良好的力学和屏蔽性能相结合的柔性中子屏蔽材料是未来发展的趋势。     

射线与中子辐射屏蔽材料的研究进展

主要对X射线、γ射线及中子辐射屏蔽材料的种类、基质材料和屏蔽性能进行了全面的分析和综述。在此基础上，探讨了现有射线及中子辐射屏蔽材料存在的主要问题，如吸收效果不高、基质材料力学性能差、屏蔽性能与其他性能难以兼顾等。最后展望了辐射屏蔽材料的研究方向，指出开发多功能、轻质、高强的抗辐射屏蔽材料是未来的发展趋势。     

乏燃料运输和储存容器中子屏蔽材料应用及研究现状

目的 了解国内外乏燃料运输和储存容器中子屏蔽材料的类型,整理分析现有中子屏蔽材料的性能和特点,为应用于乏燃料运输和储存容器的中子屏蔽材料的研发提供一定参考。方法 综述国内外应用于乏燃料运输和储存容器中子屏蔽材料的应用现状,对关键性能进行总结和比较,并提出其研究重点和发展趋势。结果 目前,硼化不锈钢、碳化硼/铝复合材料、硼铝合金、聚合物基复合材料和屏蔽混凝土等中子屏蔽材料已应用于乏燃料运输和储存容器。结论 随着核电厂高燃耗的发展趋势,未来乏燃料运输和储存容器对中子屏蔽材料的性能提出了更严格的要求,建议注重研发屏蔽性能优异、装配更换方便、耐辐照的中子屏蔽材料。     

金属氢化物中子屏蔽应用研究现状

中子屏蔽应用中,屏蔽材料的氢密度水平是影响中子屏蔽性能的关键因素;一些含氢密度高的金属氢化物具有中子屏蔽能力强、耐高温的特点,与传统材料相比可以显著减少中子屏蔽体质量和体积,是小型模块化紧凑式核反应堆、空间核反应堆应用中的重要候选屏蔽材料。本文通过文献调研,对氢化锂、氢化锆、氢化钛材料的中子辐射屏蔽性能、辐照稳定性的研究现状进行了综述。指出辐照条件下金属氢化物的辐照肿胀现象是影响其用作中子屏蔽材料的关键,总结分析了导致氢化锂、氢化锆辐照肿胀的因素,为未来金属氢化物用于中子屏蔽的研究提供参考。     

NBR/PVC橡塑合金辐射防护材料的制备及性能研究

为了满足辐射防护的要求以及发展结构功能一体化柔性防护材料的需求,研制了一种新型的橡塑合金复合防护材料。通过力学性能和屏蔽性能测试发现,该复合材料具有较高的拉伸强度与断裂伸长率,且材料的中子与伽马屏蔽性能良好,随着功能填料含量的不断增加,中子和伽马屏蔽性能也不断提升,但材料的拉伸强度和断裂伸长率不断减小。辐照实验显示,电子束硫化后的样品比硫磺硫化后的样品拥有更出色的力学性能。综合考虑该材料良好的中子与伽马屏蔽性能、拉伸性能与柔韧性能等特性,其在形状较为复杂的核动力设施以及辐射防护服等领域具有较大的应用潜力。     

新型柔性硼基苯基硅橡胶中子屏蔽材料的制备和性能研究

研制了一种新型柔性硼基苯基硅橡胶中子屏蔽材料,重点研究了不同B4C含量对复合材料的力学性能、热学性能及中子屏蔽性能的影响。结果表明:复合材料的断裂拉伸应变和拉伸强度随着B元素含量的增加有短暂的减小;复合材料的耐热性能随着B元素含量的增加而增强;复合材料的中子屏蔽性能随着B元素含量的增加而增加;纳米复合材料相比于微米复合材料在力学性能、热学性能和中子屏蔽性能方面都有一定的提高。该新型苯基硅橡胶柔性中子屏蔽材料具备应用到结构复杂、形状各异的核技术场地中的潜力,有很大的应用前景。     

核防护用水泥基中子屏蔽材料的研究进展

不同能量的中子有不同的工程屏蔽方法,水泥基中子屏蔽材料具有重要应用价值.本文首先从中子防护的角度简要介绍了中子屏蔽原理,其次从快中子减速、慢中子吸收两个方面总结概括了水泥基中子屏蔽材料的研究现状,分析了水泥基中子屏蔽材料存在的不足:功能单一、耐久性、施工性及环境友好等问题,并指出了下一步研究方向:提高核防护水泥混凝土综...     

粘结碳化硼材料中子屏蔽性能及制备工艺研究

中子准直器在中国散裂中子源(CSNS)的谱仪上起着限制束流截面与发散度的作用,准直器的结构设计中通过在中子飞行管上间隔设置B4C阻挡块可以起到阻挡、吸收屏蔽杂散中子的作用,为此研制了粘结碳化硼中子阻挡块,并对其成型工艺进行了研究,确定了成型最佳粉体尺寸、粘结剂含量和固化温度.对该中子屏蔽材料的中子衰减系数计算结果表明,波长为0.1×10-10、1×10-10和15×10-10m中子通过该材料的线性衰减系数分别为3.42、30.4和449.9 cm-1.力学冲击和热重实验结果也表明该材料符合在CSNS谱仪中子束线屏蔽使用中的性能要求.     

橄榄岩防辐射混凝土的试验研究

在辐射屏蔽混凝土中氢元素含量对于中子屏蔽具有重要作用,水中含有较大比例的氢元素,因此材料中的结晶水含量在辐射屏蔽混凝土制造中是一个关键因素。以富含结晶水(质量分数m6%)的橄榄岩为混凝土的粗、细集料,完成了混凝土配合比试验,配制了C30等级的防辐射混凝土。中子和γ射线屏蔽试验结果表明,同普通混凝土相比,橄榄岩混凝土中子辐射屏蔽性能明显增强,γ射线屏蔽性能也比较优异。     

Radiations in space: risk estimates

The complexity of radiation environments in space makes estimation of risks more difficult than for the protection of terrestrial populations. In deep space the duration of the mission, position in the solar cycle, number and size of solar particle events (SPE) and the spacecraft shielding are the major determinants of risk. In low-earth orbit missions there are the added factors of altitude and orbital inclination. Different radiation qualities such as protons and heavy ions and secondary radiations inside the spacecraft such as neutrons of various energies, have to be considered. Radiation dose rates in space are low except for short periods during very large SPEs. Risk estimation for space activities is based on the human experience of exposure to gamma rays and to a lesser extent X rays. The doses of protons, heavy ions and neutrons are adjusted to take into account the relative biological effectiveness (RBE) of the different radiation types and thus derive equivalent doses. RBE values and factors to adjust for the effect of dose rate have to be obtained from experimental data. The influence of age and gender on the cancer risk is estimated from the data from atomic bomb survivors. Because of the large number of variables the uncertainities in the probability of the effects are large. Information needed to improve the risk estimates includes: (1) risk of cancer induction by protons, heavy ions and neutrons: (2) influence of dose rate and protraction, particularly on potential tissue effects such as reduced fertility and cataracts: and (3) possible effects of heavy ions on the central nervous system. Risk cannot be eliminated and thus there must be a consensus on what level of risk is acceptable.     

Neutron threshold activation detectors (TAD) for the detection of fissions

Prompt fission neutrons are one of the strongest signatures of the fission process. Depending on the fission inducing radiation, their average number ranges from 2.5 to 4 neutrons per fission. They are more energetic and abundant, by about 2 orders of magnitude, than the delayed neutrons (≈3 vs. ≈0.01) that are commonly used as indicators for the presence of fissionable materials.The detection of fission prompt neutrons, however, has to be done in the presence of extremely intense probing radiation that stimulated them. During irradiation, the fission stimulation radiation, X-rays or neutrons, overwhelms the neutron detectors and temporarily incapacitate them. Consequently, by the time the detectors recover from the source radiation, fission prompt neutrons are no longer emitted. In order to measure the prompt fission signatures under these circumstances, special measures are usually taken with the detectors such as heavy shielding with collimation, use of inefficient geometries, high pulse height bias and gamma-neutron separation via pulse-shape discrimination with an appropriate organic scintillator. These attempts to shield the detector from the flash of radiation result in a major loss of sensitivity. It can lead to a complete inability to detect the fission prompt neutrons.In order to overcome the blinding induced background from the source radiation, the detection of prompt fission neutrons needs to occur long after the fission event and after the detector has fully recovered from the source overload. A new approach to achieve this is to detect the delayed activation induced by the fission neutrons. The approach demonstrates a good sensitivity in adverse overload situations (gamma and neutron “flash”) where fission prompt neutrons could normally not be detected.The new approach achieves the required temporal separation between the detection of prompt neutrons and the detector overload by the neutron activation of the detector material. The technique, called Threshold Activation Detection (TAD), is to utilize appropriate substances that can be selectively activated by the fission neutrons and not by the source radiation and then measure the radioactively decaying activation products (typically beta and gamma rays) well after the source pulse. The activation material should possess certain properties: a suitable half-life of the order of seconds; an energy threshold below which the numerous source neutrons will not activate it (e.g., 3 MeV); easily detectable activation products (typically >1 MeV beta and gamma rays) and have a usable cross-section for the selected reaction. Ideally the substance would be a part of the scintillator.There are several good material candidates for the TAD, including fluorine, which is a major constituent of available scintillators such as BaF₂, CaF₂ and hydrogen free liquid fluorocarbon. Thus the fluorine activation products, in particular the beta particles, can be measured with a very high efficiency in the detector.The principles, applications and experimental results obtained with the fluorine based TAD are discussed.     

Cosmic radiation shielding tests for UHMWPE fiber/nano-epoxy composites

Cosmic radiation shielding properties are important for spacecraft, and hydrogenous materials such as polyethylene have been shown to be effective in shielding against galactic cosmic rays and solar energetic particles. Ultrahigh molecular weight polyethylene (UHMWPE) fibers, which are effective in such shielding, also have advanced mechanical and physical properties, which potentially are very valuable for NASA space missions both as a radiation shield and as vehicle structure. In our previous studies, we fabricated a nano-epoxy matrix with reactive graphitic nanofibers that showed enhanced mechanical (including strength, modulus and toughness) and thermal properties (higher T_g, stable CTE, and higher ageing resistance), as well as wetting and adhesion ability to UHMWPE fibers. In this work, the radiation shielding performance of the UHMWPE fiber reinforced nano-epoxy composite was characterized by radiation tests at the NASA Space Radiation Laboratory at Brookhaven National Laboratory. The results showed that the high radiation shielding performance associated with UHMWPE was not degraded by the addition of graphitic nanofibers in the matrix. Together with the previous studies showing higher mechanical properties, these new studies validate the importance of the UHMWPE fiber/nano-epoxy composite for potential applications in more durable space composites and structures, and offer reduced manufacturing costs and wider design applications through avoidance of specialized and in some cases ineffective UHMWPE fiber surface treatment processes.     

A combined neutron and gamma-ray multiplicity counter based on liquid scintillation detectors

Multiplicity counters for neutron assay have been extensively used in materials control and accountability for nonproliferation and nuclear safeguards. Typically, neutron coincidence counters are utilized in these fields. In this work, we present a measurement system that makes use not only of neutron (n) multiplicity counting but also of gamma-ray (γ) multiplicity counting and the combined higher-order multiples containing both neutrons and gamma rays. The benefit of this approach is in using both particle types available from the sample, leading to a reduction in measurement times compared with single-particle measurements. We present measurement results of n, γ, nn, nγ, γγ, nnn, nnγ, nγγ and γγγ multiples emitted by Mixed-Oxide (MOX) samples measured at Idaho National Laboratory (INL). The MOX measurement is compared to initial validation of the detection system done using a ²⁵²Cf source. The dual radiation measuring system proposed here uses extra measurables to improve the statistics when compared to a neutron-only system and allows for extended analysis and interpretation of sample parameters. New challenges such as the effect of very high intrinsic gamma-ray sources in the case of MOX samples are discussed. Successful measurements of multiple rates can be performed also when using high-Z shielding.     

An airport cargo inspection system based on X-ray and thermal neutron analysis (TNA)

A cargo inspection system incorporating a high-resolution X-ray imaging system with a material-specific detection system based on Ancore Corporation's patented thermal neutron analysis (TNA) technology can detect bulk quantities of explosives and drugs concealed in trucks or cargo containers. The TNA process utilises a 252Cf neutron source surrounded by a moderator. The neutron interactions with the inspected object result in strong and unique gamma-ray signals from nitrogen, which is a key ingredient in modern high explosives, and from chlorinated drugs. The TNA computer analyses the gamma-ray signals and automatically determines the presence of explosives or drugs. The radiation source terms and shielding design of the facility are described. For the X-ray generator, the primary beam, leakage radiation, and scattered primary and leakage radiation were considered. For the TNA, the primary neutrons and tunnel scattered neutrons as well as the neutron-capture gamma rays were considered.     

Shielding design studies for a neutron irradiator system based on a 252Cf source.

This study aims to investigate a shielding design against neutrons and gamma rays from a source of 252Cf, using Monte Carlo simulation. The shielding materials studied were borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP4B was used to design shielding for 252Cf based neutron irradiator systems. By normalising the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independent of the intensity of the actual 252Cf source. The results show that the total dose equivalent rates were reduced significantly by the shielding system optimisation.     

High-performance heavy concrete as a multi-purpose shield

Concrete has long been used as a shield against high-energy photons and neutrons. In this study, colemanite and galena minerals (CoGa) were used for the production of an economical high-performance heavy concrete. To measure the gamma radiation attenuation of the CoGa concrete samples, they were exposed to a narrow beam of gamma rays emitted from a (60)Co radiotherapy unit. An Am-Be neutron source was used for assessing the shielding properties of the samples against neutrons. The compression strengths of both types of concrete mixes (CoGa and reference concrete) were investigated. The range of the densities of the heavy concrete samples was 4100-4650 kg m(-3), whereas it was 2300-2600 kg m(-3) in the ordinary concrete reference samples. The half-value layer of the CoGa concrete samples for (60)Co gamma rays was 2.49 cm; much less than that of ordinary concrete (6.0 cm). Moreover, CoGa concrete samples had a 10 % greater neutron absorption compared with reference concrete.     

MYRRHA, a Pb-Bi experimental ADS: specific approach to radiation protection aspects

Since 1998, SCK*CEN, in partnership with IBA s.a. and many European research laboratories, is designing a multipurpose accelerator driven system (ADS) for Research and Development (R&D) applications-MYRRHA-and is conducting an associated R&D support programme. MYRRHA is an ADS under development at Mol in Belgium and is aiming to serve as a basis for the European experimental ADS to provide protons and neutrons for various R&D applications. It consists of a proton accelerator delivering a 350 MeV x 5 mA proton beam to a liquid Pb-Bi spallation target that in turn couples to a Pb-Bi cooled, subcritical fast core. In the first stage, the project focuses mainly on demonstration of the ADS concept, safety research on sub-critical systems and nuclear waste transmutation studies. In a later stage, the device will also be dedicated to research on structural materials, nuclear fuel, liquid metal technology and associated aspects, and on sub-critical reactor physics. Subsequently, it will be used for research on applications such as radioisotope production. A first preliminary conceptual design file of MYRRHA was completed by the end of 2001 and has been reviewed by an International Technical Guidance Committee, which concluded that there are no show stoppers in the project and even though some topics such as the safety studies and the fuel qualification need to be addressed more deeply before concluding it. In this paper, we are reporting on the state-of-the art of the MYRRHA project at the beginning of 2004 and in particular on the radiation shielding assessment and the radiation protection particular aspects through a remote handling operation approach in order to minimise the personnel exposure to radiation.