炉内成球法制备SiC空心陶瓷微球

采用干凝胶法,以聚碳硅烷（PCS）为原料,通过炉内成球技术制备了SiC空心陶瓷微球。并利用TG、IR、SEM、XRD等方法对陶瓷微球进行了成键结构、表面形貌等分析,讨论了有机聚合物的陶瓷化过程机理。结果表明,干凝胶成球技术能利用经纯化处理的聚碳硅烷在500~600 ℃下得到SiC空心陶瓷微球,采用乙醇作为发泡剂可使PCS凝胶粒子得到良好发泡效果,提高载气中氦气含量至50%~80%可提高干凝胶粒子在吸热阶段的升温速率,微球经辐照后在850 ℃下碳化生成以β-SiC为主要相结构的球壳,球壳具有较好的表面平整度。     

预处理温度对聚碳硅烷粒子炉内成球性的影响

本文以聚碳硅烷(PCS)为原料,采用炉内成球技术制备直径200~400μm、壁厚3~5μm的SiC空心微球,探讨微球制备的最佳条件,并在此基础上研究不同预处理温度对PCS成球产率及品质的影响。结果表明,炉内载气温度为500℃、He与Ar比例为3∶1时PCS的成球产率较高,且微球的球形度、同心度、表面光洁度均最好。此外,由于预处理过程去除了PCS中的低分子量聚碳硅烷和其他小分子,同时使其聚合度升高,提高了PCS的热稳定性和陶瓷化产率。因此,在最佳炉内成球条件下,PCS的成球率随预处理温度的升高而升高,所得微球的表面粗糙度却随之降低。经350℃预处理后的PCS粒子成球率最高,且微球的球形度和表面质量最佳。     

高温气冷堆新燃料元件运输容器临界安全分析

采用基于蒙特卡罗方法的MCNP5程序对高温气冷堆所用的球形燃料元件进行描述;根据包覆燃料颗粒在燃料球内的分布性质构建了8种不同模型,并研究不同模型对有效增殖因子（keff）和计算时间的影响,获得了临界计算问题中最优的燃料球模型;运用MCNP5描述燃料球运输容器,并研究了容器中子吸收板厚度、外容器壁厚、缓冲层材料、反射层材料、容器形状、容器结构缺失和水密度等影响运输容器临界安全的因素。结果表明,所研究的高温气冷堆新燃料元件运输容器在正常运输条件下和事故运输条件下均处于临界安全状态,其临界安全指数（CSI）可定为0。      

塑料微球强度和渗氘性能测量

摘要：针对激光惯性约束聚变物理实验打靶所需的氘氚燃料塑料微球靶进行微球强度和渗氘性能初步研究,测量了塑料微球球壳在20（室温）、40、60℃下的压缩强度,并实验测得了塑料微球在20、60℃下的氘气渗透率。     

包覆工艺对微球形燃料相弥散燃料混合均匀性的影响

针对微球形燃料相颗粒与基体粉末的流动性相差较大、难于混合均匀,建立了一种微球的包覆工艺,并研究了包覆工艺对混合均匀性的影响。采用直径约为100μm的不锈钢微球代替燃料微球,研究结果表明,在微球表面物理包覆一层基体粉末,可增加颗粒表面粗糙度,降低两组元粉末的密度差及颗粒沉降的距离,包覆层还能使颗粒间保持一定的间距,微观上形成连续的基体网络,减少甚至避免发生偏聚,有效地改善了混合均匀性。包覆工艺的最佳参数为:保温温度,76℃;保温时间,6min;黏结剂添加量,1%;粉末粒径,小于25μm。该方法可用于改善(U-Mo)-Al、(U-Mo)-Zr等微球形燃料相弥散燃料的混合均匀性。     

U-M_o微球物理包覆工艺研究

为了减少U—M_o微球在混料和成型过程中发生偏聚的现象,改善(U-M_o)-Al弥散燃料铀分布的均匀性,研究了多种U-M_o微球物理包覆工艺及建立的包覆工艺对微球流动性的影响。本文采用相近尺寸(直径约100μm)的不锈钢微球代替U-M_o微球。结果表明:直接添加黏结介质包覆工艺和自由落体包覆工艺不适合微球的包覆;采用溶剂挥发法在每颗微球表面都包覆了一层基体粉末颗粒,微球表面的粗糙度明显增大,流动时颗粒间产生机械咬合力,降低了微球的流动性。     

空气中两种辐照场下聚碳硅烷先驱丝的辐照不熔化效应

采用⁶⁰Co产生的γ射线与10MeV电子加速器产生的电子束在空气中辐照聚碳硅烷先驱丝。利用凝胶含量测定、红外光谱分析、凝胶渗透色谱分析、热重分析等手段研究了经两种辐照场辐照后聚碳硅烷先驱丝的凝胶含量、化学结构、分子量分布及其热分解特性的变化。结果表明：两种辐照场下能满足聚碳硅烷先驱丝不熔化处理的剂量分别为3、6.5MGy,两种辐照场下聚碳硅烷先驱丝的辐照不熔化机理一致,均通过形成Si—C—Si、Si—O—Si桥联结构而实现不熔化处理;在相同吸收剂量下,γ射线辐照不熔化效果较电子束辐照的好;γ射线辐照的聚碳硅烷先驱丝陶瓷产率也较电子束辐照的高。     

UO_2-Zr弥散燃料板氧化扩散机理研究

UO2-Zr弥散燃料板的氧化过程包括包壳与冷却剂的氧化反应和芯体中弥散的UO2燃料微球氧原子扩散过程。本文通过直接求解球坐标系下的氧化扩散方程,得到UO2燃料微球高温下向芯体中氧原子扩散强度的解析式,该式与实验数据符合良好,并结合锆水反应与UO2燃料微球高温氧原子扩散效应构建了UO2-Zr板的氧化扩散模型。新模型能预测不同的氧化结构、芯体中更高的氧原子浓度以及相对较低的氧化吸氧量,为UO2-Zr板严重事故早期行为的研究提供了理论基础。     

UO2-(Zr0.8Ca0.2)O1.8面心立方固溶体微球制备工艺研究

为研制出耐辐照的新型单相陶瓷燃料,采用溶胶-凝胶法,通过复合溶胶配制、分散胶凝、洗涤、干燥煅烧与烧结过程,开展了UO2-(Zr0.8Ca0.2)O1.8燃料微球制备工艺研究,制备出铀摩尔分数含量分别为30mol%、50mol%、70mol%的UO2-(Zr0.8Ca0.2)O1.8燃料微球样品。在对工艺过程进行分析的基础上,通过实验确定了工艺参数。采用X射线衍射（XRD）对3种燃料微球样品进行分析,分析结果表明:铀摩尔分数含量分别为30mol%、50mol%、70mol%的UO2-(Zr0.8Ca0.2)O1.8燃料微球样品均为面心立方（FCC）固溶体结构。      

抗溶剂气体重结晶制备聚碳硅烷纳米微粒的溶剂选择

将理论计算和实验相结合来选择聚碳硅烷(PCS)抗溶剂气体重结晶过程的溶剂。首先，理论计算了聚碳硅烷的溶度参数，并根据极性相似相溶原理筛选出几种待选溶剂；然后，实验比较CO2对待选溶剂的体积膨胀，待选溶剂与CO2、PCS的溶解情况和PCS在待选溶剂中的析晶率。研究结果表明：在3种待选溶剂四氢呋喃、对二甲苯和环己烷中，以四氢呋喃的性能最佳。以四氢呋喃作溶剂，采用抗溶剂气体重结晶工艺制备出的PCS纳米微粒平均粒径为2．2nm，颗粒粒径分布较为集中。     

惯性约束聚变用聚酰亚胺靶丸的研究进展

综述了近年来国内外在惯性约束聚变实验用聚酰亚胺(PI)靶丸方面的研究与应用进展情况.从ICF实验用靶的性能要求、常用PI的化学结构、PI靶丸的装配方法以及目前PI靶丸应用过程中存在的问题等几个方面进行了论述.在此基础上提出了PI靶丸的研究方向.     

废放射源整备用封装管的设计与检测

研制废放射源整备封装管,使整备后废放射源满足处置或者长期贮存要求,是废放射源管理的重要组成部分。本文针对不同活度、不同核素的废放射源,设计制造了不同的封装管,并对其中的螺纹封装管进行了跌落性、抗冲击性、耐热性等一系列检测。检测结果表明,本研究设计的封装管,满足封装废放射源的要求。     

Observations of a steady state effect limiting radiation damage in reactor vessel steels

Reactor vessel material surveillance capsules which contain specimens of actual material used in the construction of a vessel are contained in nearly all operating reactors. These specimens monitor the changes in properties of the reactor vessel and assure that predicted changes based on trend curves which are used to set operating limits for the plant are conservative. Recently, data has been obtained from the Point Beach Unit No. 1 and Connecticut Yankee reactor vessel surveillance capsules exposed to neutron radiation for much longer periods of time, than those irradiated in test reactors and surveillance capsules which were removed at the first refueling and other early refueling outages. The data from these long time surveillance capsule exposures when compared to data from capsules removed from the same reactors earlier in life indicated that a limiting or steady state condition has resulted rather than a continuous embrittlement as predicted by trend curves. It is believed that the limited embrittlement or steady state condition which occurred from the surveillance capsule tests is due to a combination of relatively low neutron flux compared to that existing in test reactors which were the primary source of data used to establish trend curves and the longer exposure periods in the capsules that led to significant “annealing” during irradiation.     

内表面掺硫聚苯乙烯空心微球初步研制

以聚苯乙烯磺酸钠（PSS）为掺杂源,采用乳液微封装法制备内表面掺S的聚苯乙烯（PS）空心微球。XPS测量表明,微球内表面存在明显的S元素。利用低能X射线照相技术研究了不同掺杂源浓度下干燥时重力和表面张力对掺杂均匀性的影响,当掺杂源浓度降为0.2%时,得到了内表面掺杂S均匀的PS微球,微球直径为200～800μm,直径300μm的微球中掺杂层S原子面密度为8.57×10⁶μm^－2。     

Conversion of organic-inorganic polymers to ceramics by ion implantation

Thin films of polycarbosilane (PCS) and polysiloxane (SR350 resin) deposited on Si substrates were irradiated with He, C, or Au ions in order to convert them into SiC and SiOC ceramic coatings. The transformation kinetics was assessed by means of ion beam analysis (RBS, NRA, ERDA) and of FTIR, while the hardening was measured by nanoindentation tests. The yield of H release increased exponentially with the amount of electronic excitations up to a saturation value, and correlated properties of compaction and hardness varied in proportion to this yield (SR350) or to its square (PCS). PCS films became as hard as amorphous SiC and SR350 films 60% harder than silica. A relative softening was observed in the case of Au irradiations with respect to lighter ions, due to the atomic disordering by collision cascades. The irradiations improved also the oxidation resistance during subsequent annealings.     

堆用热管的热特性计算分析

热管已广泛地用于许多领域。为把热管技术用于材料试验堆内的辐照罐温度控制,近年来日本原子力研究所 JMTR 堆上的研究人员对热管作了大量的研究工作,并进行了堆外热特性试验和部分堆内试验。作者在 JMTR 堆上工作期间,主要对几种堆用垂直重力协助式热管的热特性进行了计算分析。研究结果表明,轴向槽道式热管和无吸液热管不仅结构简单,而且具有良好的热特性,可工作在500K 左右的温度范围内,适用于考验压水堆材料的辐照罐。均匀吸液芯热管在较高温度下沸腾极限较低,因此更适合于较低温度下的使用。本文简介热管在辐照技术上的应用并讨论几种堆用重力热管的传热极限。     

Twin Astir: An irradiation experiment in liquid Pb-Bi eutectic environment

The Twin Astir irradiation program, currently under irradiation in the BR2 reactor at SCK.CEN is aimed at determining the separate and possibly synergetic effects of a liquid lead bismuth eutectic (LBE) environment and neutron irradiation. It will lead to a parameterisation of the key influencing factors on the mechanical properties of the candidate structural materials for the future experimental accelerator driven system (ADS). The experiment consists of six capsules containing mainly mini tensile samples and one capsule containing mini DCT’s (disc shaped compact tension specimens). Three of the tensile containing capsules and half of the DCT containing capsule are filled each with approximately 20 ml of low oxygen (10⁻⁶ wt%) LBE. To complete the filling of these capsules with LBE under controlled conditions a dedicated filling installation was constructed at SCK.CEN. The other three tensile containing capsules are subjected to PWR water conditions, in order to discriminate the effect of PbBi under irradiation from the effect of the irradiation itself. To extract the effect of the PbBi corrosion itself on the material properties, one of the capsules is undergoing the thermal cycles of the BR2 reactor without being subjected to irradiation. This results in a matrix of three irradiation doses in LBE (0, 1.5 and 2.5 dpa) and two environments (PbBi and PWR water conditions). Here we will present the detailed concept and the status of the Twin Astir project, describe the materials under irradiation and report on our experience with the licensing of the experiment.