裂变靶辐照装置水力特性试验技术

针对裂变靶辐照装置内外流道尺寸狭小的结构特点及实堆测量需求,提出了进行裂变靶辐照装置水力特性试验的技术方案,采用微小压力传感器实测狭窄套管内外流道的流动压差,通过单项标定试验获得内外流道压差-流量关系拟合公式.由该拟合公式得到的计算值与内外管实测流量的相对误差最大不超过士2%,可以用于裂变靶辐照装置在实堆中进行水力模拟的试验数据处理.     

裂变同位素靶管在300#反应堆的辐照实验

将裂变同位素靶管放入300#反应堆特殊燃料元件盒中辐照.叙述了辐照定位装置的加工以及靶管取放、定位、热电偶导出线固定等详情,测量了轴向中子注量率分布、水隙内和靶管内外壁的中子注量率,刻度了靶管的反应性,分五个功率台阶测量了靶管壁面温度和堆池温度.测量得到的热中子注量率沉降因子与根据生产厂家报告推出的自屏因子基本符合.将计算的和测量的靶管外壁温度进行比较,可认为铀靶生产厂家给出的靶管各热物性参数是准确的.     

钼锝靶件辐照装置堆外自然循环实验研究

钼锝同位素辐照装置的堆外热工水力研究,对于确定靶件的辐照参数和辐照安全具有重要意义,相关结果可为靶件的设计验证和入堆辐照提供技术支持。本文针对自主研发的钼得靶件辐照装置,设计了堆外传热实验台架,开展了钼锝靶件辐照装置堆外传热验证实验,对比分析了不同热流密度、不同模拟靶件外径、不同辐照装置结构等条件下实验段的温度分布。结果表明,辐照装置内导流管的设计可有效提高装置的自然循环能力,证明了所设计辐照装置的安全性。     

裂变快中子辐照在GaAs中产生缺陷的正电子湮没研究

采用自电子湮没寿命测量方法研究了注量为６．５×１０＾１５／ｃｍ＾２和１．４×１０＾１４／ｃｍ＾２，Ｅｎ≥１ＭｅＶ的裂变中子辐照在掺Ｓｉ，Ｎ型单晶ＧａＡｓ产物的缺陷，此辐照在ＧａＡｓ中产生单空位和双空位缺限，缺陷浓度于比于辐照注量，高温退火产生三空位缺陷及小空位团，单空位，双空位和三空位缺陷的退火温度分别为２５０，４５０，６５０℃。     

堆内辐照过程中辐照靶件的核发热和传热研究

研究了辐照靶件在堆内辐照过程中的核发热和传热计算方法。温度的计算结果与实验结果符合良好。这表明：核发热和传热计算方法可为放射性同位素辐照生产和辐照安全提供重要参数和保证。     

MHWRR在极低光激中子水平下的临界启动技术

阿尔及利亚比林核研究中心重水反应堆(Multi-purposes Heavy Water Research Reactor,MHWRR)实施了仪表、控制和电气数字化升级改造,改造后的首次临界启动对改造工程具有重要意义。为保证反应堆启动安全,需要解决升级改造后在极低光激中子水平下临界启动中存在的核测量盲区问题,首先对长期停堆后堆内剩余光激中子源强、核测量盲区以及临界棒位进行了理论计算与分析研究,在此基础上提出了在无外加启动中子源条件下首次临界启动的实验技术方案。在无参考数据的情况下,实验进程完全按理论设计的预期进行,临界启动一次成功;启动过程中核功率参数得到有效监测,启动测量装置与堆外电离室测量范围衔接完备,临界棒位理论值与实验值的误差小于0.84%,实验结果与理论计算结果符合良好,表明了这项实验技术的合理可行。     

高浓铀靶裂变法生产钼-99发展的综述

随着放射性药物化学和核医学的快速发展,放射性诊断核素^99mTc在临床的应用越来越广泛,从而使得当前全球对其母体核素⁹⁹Mo的需求量不断增加。目前高浓铀靶裂变法生产⁹⁹Mo仍是最广泛使用的方法。本文系统介绍了国内外采用高浓铀靶裂变法生产⁹⁹Mo的发展历史和生产工艺。     

CARR水冷同位素生产研究

水冷同位素的生产是CARR堆的一项重要应用,通过对CARR生产水冷同位素的可行性研究,提出了一系列方案并获得初步结果。从经济性和安全性角度对靶件进行的设计优化计算工作如下。 1)水冷同位素产额计算分析。经济性指标计算结果的准确性对于方案优化好坏有重大影响。 2)应用Monte Carlo法直接得到反应率,大大提高了同位素产额和靶件释热率计算精度。     

HFETR辐照靶件设计程序开发

针对高通量工程试验堆（ＨＦＥＴＲ）燃料材料考验的辐照靶件设计，开发了ＧＥＮＧＴＣ－Ｃ程序，可用它进行辐照靶件设计和温度分布计算，程序可对多层包壳材料传热，间隙层（液体或气体介质）传热，气体介质层传热进行了计算，同时，对热导膜率随温度的变化和靶件结构材料的几何尺寸因热膨胀的改变进行修正，原程序由美国ＯＲＮＬ编制，见长于结构严谨流畅和良好的适用性，因此，ＧＥＮＧＴＣ－Ｃ是在原程序的基础上并结合ＨＦＥＴ     

低能氦离子辐照对钨和钼材料的表面损伤作用

钨和钼材料具有高熔点、高热导率、低溅射率等优点成为国际热核实验反应堆计划中面向等离子体材料的候选材料。因此研究钨和钼材料的辐照损伤行为对于认识面向等离子体材料的辐照损伤机制具有重要意义。本文采用120 e V的He+在873 K对钨和钼材料进行辐照实验,利用纳米压痕仪与导电模式原子力显微镜(Conductive Atomic Force Microscopy,CAFM)相结合,原位比较了钨和钼材料在辐照前后的表面形貌、表面微结构以及表层缺陷分布的变化特征。结果表明,低能He+辐照会导致钨和钼材料的近表面产生纳米量级氦泡缺陷,这些氦泡缺陷的存在使得样品表面产生绒毛或波浪状结构。纳米压痕深度分析和扫描电镜的分析结果表明,低能He+辐照会对Mo材料产生明显的刻蚀作用。本工作对于进一步认识低能氦离子辐照对面向等离子体材料的辐照损伤作用具有一定的科学参考意义。     

Low temperature neutron irradiation effects on microstructure and tensile properties of molybdenum

Polycrystalline molybdenum was irradiated in the hydraulic tube facility at the High Flux Isotope Reactor to doses ranging from 7.2 × 10⁻⁵ to 0.28 dpa at 80 °C. As-irradiated microstructure was characterized by room-temperature electrical resistivity measurements, transmission electron microscopy (TEM) and positron annihilation spectroscopy (PAS). Tensile tests were carried out between −50 and 100 °C over the strain rate range 1 × 10⁻⁵ to 1 × 10⁻² s⁻¹. Fractography was performed by scanning electron microscopy (SEM), and the deformation microstructure was examined by TEM after tensile testing. Irradiation-induced defects became visible by TEM at 0.001 dpa. Both their density and mean size increased with increasing dose. Submicroscopic three-dimensional cavities were detected by PAS even at 0.0001 dpa. The cavity density increased with increasing dose, while their mean size and size distribution was relatively insensitive to neutron dose. It is suggested that the formation of visible dislocation loops was predominantly a nucleation and growth process, while in-cascade vacancy clustering may be significant in Mo. Neutron irradiation reduced the temperature and strain rate dependence of the yield stress, leading to radiation softening in Mo at lower doses. Irradiation had practically no influence on the magnitude and the temperature and strain rate dependence of the plastic instability stress.     

Microstructure damage of titanium films by irradiation with fission fragments

Radiation damage caused by fission fragments to metal surfaces is an important research topic. Thin titanium foils were irradiated with a continuous wave beam of 132 MeV ¹³²Xe⁺²⁹ at the current intensity of 2 pnA. Pre- and post-irradiated surface topologies were investigated using atomic force microscopy and the observed defects were quantified by root mean square roughness, depth profile of the disordered zones, size and areal density of the voids, and discussed as a function of the applied fluencies (1-9) × 10¹³ Xe/cm². The first ellipsoidal dislocation loops appeared at the fluence of 3.0 × 10¹³ Xe/cm² with the areal density of 1.56 × 10⁶/cm² that increased to 2.0 × 10⁷ cm⁻² when the dose rose to 9.0 × 10¹³ Xe/cm². At this point also the first dislocation lines with the density of 1.3 × 10⁷ cm⁻² were seen. Our results suggest that the fission fragments might maximize large voids and dislocations and increase the degradation in depth resolution.     

Modeling of fission gas swelling in LWR UO2 fuel under irradiation

An understanding of the behavior of fission gas in uranium dioxide (UO₂) fuel is necessary for the prediction of the performance of fuel rods under irradiation. A mechanistic model for matrix swelling by the fission gas in LWR UO₂ fuel is presented. The model takes into account intragranular and intergranular fission gas bubbles behavior as a function of irradiation time, temperature, fission rate and burn-up. The intragranular bubbles are assumed to be nucleated along the track of fission fragments, which play the dual role of creator and destroyer of intragranular bubbles. The intergranular bubble nuclei is produced until such time that a gas atom is more likely to be captured by an existing nucleus than to meet another gas atom and form a new nucleus. The capability of this model was validated by a comparison with the measured data of fission gas behavior such as intragranular bubble size, bubble density and total fuel swelling. It was found that the calculated intragranular bubble size and density are in reasonable agreement with the measured results in a broad range of average fuel burn-ups 6–83 GW d/tU. Especially, the model correctly predicts the fuel swelling up to a burn-up of about 70 GW d/tU.