反应堆乏燃料贮运用中子吸收材料的研究进展

从我国对乏燃料贮运用中子吸收材料的需求出发,简述了乏燃料贮运用中子吸收材料的特点、国内外研究及应用现状。重点阐述了含硼不锈钢、B4C/Al、硼铝合金、含硼有机聚合物4种含硼中子吸收材料的制备工艺、性能以及存在的问题,同时对目前我国使用的不锈钢包覆金属镉中子吸收材料和国外正在研究的含钆合金中子吸收材料进行了概述。提出了B4C/Al和硼钢两种中子吸收材料应作为进一步研究的重点。     

乏燃料贮存格架在组件跌落事故中的冲击分析

为解决乏燃料贮存格架在组件跌落事故中的冲击分析问题,对ANSYSLS-DYNA程序在弹塑性条件下的冲击分析功能进行开发,并运用于300 MW压水堆核电厂乏燃料贮存格架受乏燃料组件撞击的仿真分析中,采用应变失效和变形失效的方法对乏燃料贮存格架进行安全评定,证明现有的设计满足安全要求。所应用的弹塑性冲击分析方法,以及对乏燃料贮存格架评定的方法同样适用于其它核电设备。     

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。     

用于反应堆乏燃料贮存和运输的B4C/Al复合材料研究进展

B4C/Al复合材料是一种集结构和功能于一体的中子吸收材料,在反应堆乏燃料贮存和运输领域有着广阔的应用前景.综述了B4C/Al复合材料的主要制备工艺及国内外研究现状,并展望了未来的发展方向,最后指出随着我国核电事业的发展,B4C颗粒增强铝基复合材料将作为研究重点并在辐射屏蔽领域广泛应用.     

中子毒物硼和钆在乏燃料后处理中的应用及合金相图研究

随着我国核电站在建及运行的反应堆日渐增多,乏燃料的产生、运输和储存成为核工业面临的严峻难题。对卸载出的巨量核废乏燃料,采用闭环处理循环利用,比传统深埋更加环保和安全。采用添加中子毒物的核结构材料制造相关设备和仪器,可以保障乏燃料后处理过程的运行安全。基于核工业应用要求,针对乏燃料后处理常用的316L不锈钢、Ti35、Zr等材料添加B或者Gd的相图展开调研,结果显示B和Gd可明显提高结构材料的中子吸收效果,但B和Gd的化学性质与过渡族金属Fe, Ti, Zr等相差甚远,除α-Zr最多固溶2.8%Gd(原子数分数)以外,B和Gd均不能大量固溶于常规金属合金体系中,强制添加势必对合金力学性能或耐蚀性能带来巨大危害。可采用团簇模型的合金设计理念,通过含B和Gd化合物相均匀析出或者复合的制造工艺以及Zr-Gd、Ti-Gd体系的多元合金化,引入互溶的中间相,增加难以固溶的元素的固溶度,以此解决中子毒物B和Gd在乏燃料后处理中的应用瓶颈。     

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。     

核乏燃料运输容器减震器填充材料研究进展

目的综述以木材、聚氨酯泡沫、蜂窝铝作为核乏燃料运输容器减震器填充材料的性能。方法通过对3种填充材料的减震器进行受载分析,然后对3种填充材料的力学性能进行验证。结果将3种材料作为核乏燃料减震器填充材料具有可行性,还列出了3种材料的不足之处。结论工业上对有着优秀能量吸收、限制过载、质轻且受环境变化影较小的核乏燃料运输容器减震器填充材料有着迫切需求。之后提出将2种材料(泡沫铝和双向瓦楞蜂窝铝)作为减震器的填充材料,并分析了2种材料的特点与吸能特性,证明了其作为减震器填充材料的合理性与可行性。     

热中子防护Gd/Al复合材料的设计

随着核电和核动力的发展,热中子防护材料需具备屏蔽性能好、质量轻和机械强度高等特点,而Al基复合材料特别是Al基稀土材料正具备这一特点,因此受到研究人员的广泛关注.本文针对热中子屏蔽Gd/Al复合材料,采用蒙特卡罗方法计算了材料的热中子透射率,通过对比分析得到功能粒子Gd对纯Al材料的中子透射率的改善效果要优于功能粒子B4 C的结论.当Gd/Al复合热中子屏蔽材料的厚度大于2 mm,Gd质量含量应不需大于6％.最后将本文的研究成果应用在乏燃料储存格架材料优化设计中,并得到Gd/Al复合材料辐射性能最佳配比是Gd质量含量4％、复合材料厚度3 mm.     

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...     

功能/结构一体化中子屏蔽材料的研究现状

随着反应堆设计技术的发展,功能/结构一体化屏蔽材料成为一种发展趋势,要求中子屏蔽材料不仅具备中子屏蔽功能,而且可以兼作结构材料。中子屏蔽材料采用一体化设计可大幅简化屏蔽结构,实现屏蔽结构的轻量化和小型化。简述了功能/结构一体化中子屏蔽材料的设计要求和常见的热中子吸收核素。重点阐述了硼钢和具备功能/结构一体化潜力的铝基碳化硼复合材料、含Gd不锈钢、B/Pb复合材料的研究现状及存在问题。最后指出了功能/结构一体化中子屏蔽材料的发展方向。     

Neutron measurements around storage casks containing spent fuel and vitrified high-level radioactive waste at ZWILAG

Spectrometric and dosimetric measurements were made around a cask containing spent fuel and a cask containing high-level radioactive waste at the Swiss intermediate waste and spent fuel storage facility. A Bonner sphere spectrometer, an LB 6411 neutron monitor and an Automess Szintomat 6134A were used to characterise the n-gamma fields at several locations around the two casks. The results of these measurements show that the neutron fluence spectra around the cask containing radioactive waste are harder and higher in intensity than those measured in the vicinity of the spent fuel cask. The ambient dose equivalents measured with the LB 6411 neutron monitor are in good agreement with those obtained using the Bonner spheres, except for locations with soft neutron spectra where the monitor overestimates the neutron ambient dose equivalent by almost 50%.     

Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of ²³⁵U, ²³⁸U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.     

Characterisation of mixed neutron photon workplace fields at nuclear facilities by spectrometry (energy and direction) within the EVIDOS project

Within the EC project EVIDOS, 17 different mixed neutron-photon workplace fields at nuclear facilities (boiling water reactor, pressurised water reactor, research reactor, fuel processing, storage of spent fuel) were characterised using conventional Bonner sphere spectrometry and newly developed direction spectrometers. The results of the analysis, using Bayesian parameter estimation methods and different unfolding codes, some of them especially adapted to simultaneously unfold energy and direction distributions of the neutron fluence, showed that neutron spectra differed strongly at the different places, both in energy and direction distribution. The implication of the results for the determination of reference values for radiation protection quantities (ambient dose equivalent, personal dose equivalent and effective dose) and the related uncertainties are discussed.     

Neutron spectrometry at the interim storage facility for spent nuclear fuel

Dosimetric characteristics of neutron and photon components of mixed fields around casks for spent nuclear fuel have been determined at various places at the dry interim storage facility. The results obtained with metrological grade instruments were compared with data provided by usual survey meters for both neutrons and photons.     

Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for β-Zr.Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.     

Dosimetry at an interim storage for spent nuclear fuel

The Czech nuclear power plant Dukovany started its operation in 1985. All fuel spent from 1985 up to the end of 2005 is stored at a dry interim storage, which was designed for 60 CASTOR-440/84 casks. Each of these casks can accommodate 84 fuel assemblies from VVER 440 reactors. Neutron-photon mixed fields around the casks were characterized in terms of ambient dose equivalent measured by standard area dosemeters. Except this, neutron spectra were measured by means of a Bonner sphere spectrometer, and the measured spectra were used to derive the corresponding ambient dose equivalent due to neutrons.     

Bonner sphere neutron spectrometry at spent fuel casks

For transport and interim storage of spent fuel elements from power reactors and vitrified highly active waste (HAW) from reprocessing, various types of casks are used. The radiation exposure of the personnel during transportation and storage of these casks is caused by mixed photon–neutron fields and, frequently, the neutron dose is predominant. In operational radiation protection, survey meters and even personal dosemeters with imperfect energy dependence of the dose-equivalent response are used, i.e. the fluence response of the devices does not match the fluence-to-dose equivalent conversion function. In order to achieve more accurate dosimetric information and to investigate the performance of dosemeters, spectrometric investigations of the neutron fields are necessary. Therefore, fluence spectra and dose rates were measured by means of a simple portable Bonner multisphere spectrometer (BSS). The paper describes briefly the experimental set-up and evaluation procedure. Measured spectra for different locations, types of casks and inventory are discussed. The spectra provide a basis to determine dose rates and other integral quantities with higher accuracy and for choosing suitable area monitors, respectively, to establish correction factors applied to the dosemeter reading.