弥散型燃料的裂变气体行为研究

探讨了弥散型燃料中对辐照肿胀有重要影响的裂变气体的行为机理。裂变气体原子聚集成气泡引起燃料相肿胀,气泡的尺寸分布是影响辐照肿胀的重要因素。决定气泡生长的裂变气体的行为机理主要有：裂变气体原子的产生和热扩散迁移,气泡的成核和聚合长大,气泡内气体原子的重溶,燃料相的辐照亚晶化等过程。燃料中各种尺寸的气泡浓度随时间的变化率可用气泡生长的动力学速率方程组来描述。当裂变密度较高时,辐照产生的缺陷引起燃料相的     

UO2燃料裂变气体渗流模型研究

为分析UO₂燃料晶界气泡连通导致裂变气体间歇性释放的动力学过程,从而解决目前扩散模型预测的沿芯块径向释放份额与实验测量不符的问题,采用二维渗流模型模拟UO₂燃料晶界气泡网络的演化及与燃料棒内自由空间连通的释放过程。研究结果表明,渗流模型预测沿芯块径向的裂变气体释放份额在芯块中间部分出现局部峰值,并随着时间向芯块外侧推进,与辐照试验观察到不同燃耗下径向裂变气体分布现象定性符合。因此,本研究建立的渗流模型能够从机理上解释此前扩散模型未能预测的UO₂燃料裂变气体释放份额沿径向非单调分布现象。     

燃料棒芯块中Kr、Xe的低温释放机理

在Wise反冲机理微观模型的基础上,考虑芯块倒角的影响,给出了倒角表面平均反冲效率的计算模型;击出机理参考Olander的理论;由此建立了一个更为精细的裂变气体低温释放微观模型.使用新建模型进行计算的结果显示,倒角表面的平均反冲效率约为圆柱体芯块表面的两倍;当倒角的表面积与圆柱体芯块表面积相比不可忽略时,裂变气体反冲释放份额的计算应考虑倒角的影响.     

高温堆燃料元件和包覆颗粒对裂变产物的滞留性能研究

高温气冷堆包覆燃料颗粒的包覆层是阻止放射性裂变产物释放的第一道屏障。本文在简介ＦＲＥＳＣＯ模型的基础上,采用ＦＲＥＳＣＯ２程序,计算了＾１３７Ｃｓ,＾９０Ｓｒ,＾１１０ｍＡｇ从包覆颗粒和燃料元件的释放份额,分析比较了包覆层和石墨基体对这几种核素的滞留性能,肯定了ＴＲＩＳＯ包覆层对金属裂变产物的滞留作用。     

裂变径迹热年代学的新进展

介绍了裂变径迹热年代学的现状、最新进展以及中国科学院核分析技术开放实验室在研究和应用方面的成果．     

MEASUREMENT OF PROMPT NEUTRON SPECTRA OF ^238U FISSION INDUCED BY 10.17 AND 12.12 MeV NEUTRONS

Experimental method to measure the prompt neutron spectra of 238U fission induced by fast neutrons has been developed at HI-13 Tandem Van de Graaff Accelerator Laboratory of CIAE.These techniques employ a multi-segment fission chamber and two liquid scintillator neutron detectors.TOF(time of flight)techniques are used for primary neutrons to select the fission events induced by monoenergetic neutron from 2H(d,n) reactions instead of breakup neutrons from 2H(d,np) reactions.The fission neutron TOF spectra are measured in coincidence with the fission fragments to distinguish fission neutrons from other secondary neutrons.The method permits measurements to a fairly good accuracy under large neutron and gamma ray background.The techniques are described and experimental spectra are presented.     

裂变热电偶的研制

介绍了裂变热电偶研制中一些问题的考虑．以及裂变热电偶的初步性能测试。裂变热电偶中裂变珠的直径为１ｍｍ左右．探测热中子的灵敏度约为３×１０９ｎ／ｃｍ２·℃．时问响应好于１μｓ。裂变热电偶最明显的应用是描绘堆内和堆周围的中了通量,可作为反应堆诊断学和反应堆控制方面的新工具。     

基于WIMS-D库燃耗数据的更新与基准验证

随着AP1000等新一代压水堆的发展,燃耗深度在不断提高,平均卸料燃耗深度提高到60 GW d·t-1。然而,传统使用的WIMS69群和XMAS172群WIMS-D格式多群常数库,其能群结构存在共振峰重叠,核素种类较少,裂变产物产额的偏差较大,并且伪裂变产物中包含的核素种类较多而导致152Gd、160Gd、159Tb、160Tb等重要核素无法得到精确处理等问题。因此,本文主要针对AP1000等新一代反应堆的设计以及运行特点,基于ENDF/B-VII.1库,并且在现有基础上针对WIMS-D库中的伪裂变产物、裂变产物燃耗链以及裂变产物产额等燃耗数据进行更新,再通过NJOY程序开发了SHEM281群WIMS-D格式多群常数库。通过DRAGON程序挂载该WIMSD281库,对其进行临界和燃耗两方面基准验证。计算结果表明,该数据库的计算结果与基准值符合较好,精度较高,结果可靠,可初步用于压水堆的相关计算。     

~(252)Cf自发裂变中子诱发~(235)U裂变~(99)Mo累计产额测量

一、引言 裂变谱中子诱发~(235)U裂变~(99)Mo累计产额一直存在较大分歧。首先,苏联在五十年代测量的两个结果就相差8％。以后的测量结果虽然差距减小,但仍分为两组,彼此相差5％;一组接近苏联的高值6.4,另一组接近并稍低於热中子诱发~(235)U裂变的~(99)Mo累计产额6.14。在裂变产物产额的测量中,~(99)Mo往往被取作参考核,其产额数据的上述分歧,影响其他裂变产物产额的取值。因此,我们对其进行了测量。     

考虑扩散动力学过程的裂变道理论

考虑扩散动力学过程的裂变道理论王书暖（中国原子能科学研究院，中国核数据中心北京，１０２４１３）考虑激发核从形变基态到鞍点的集体形变运动经历着非平衡统计过程。由于核形变过程的驰豫时间远大于内部自由度驰豫时间，因此可以假定激发核鞍点以内的单粒子运动自由度...     

钍-铀燃料后处理中新萃取剂的研究——Ⅳ.N-正辛基己内酰胺对主要裂片元素的萃取及γ辐照对萃取行为的影响

研究了N-正辛基己内酰胺(OCLA)从硝酸介质中对主要裂片元素Zr、Nb、Ru及Cs的萃取性质并和相同条件下的TBP萃取性质作比较。结果表明,OCLA的萃取性质类似于TBP,但在相同条件下,OCLA萃取主要裂片元素的分配系数要比TBP高。还研究了辐照对OCLA及TBP萃取性质的影响。OCLA的辐照稳定性比TBP好,特别是辐照对萃取锆的影响比TBP萃取锆的影响小得多。     

固体径迹探测器测定无限薄裂变源的有效原子数

本文叙述了无限薄裂变源有效原子数的定义及用固体径迹探测器测量有效原子数的原理和方法,并对测量值进行了实验验证,得到了比较好的结果。     

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。     

公众吸入混合裂变产物后内照射剂量的简便估算方法(以ICRP 1994年呼吸道模型为依据)

作者前曾参照国际放射防护委员会第３０号出版物（ＩＣＲＰ－３０）给出了摄入混合裂变产物后内照射剂量的简便估算方法。现根据ＩＣＲＰ１９９４年人类呼吸道模型,重新给出了这一估算方法。计算出衰变速率系数ｎ＝０．２～２．０时呼吸道与消化道靶组织所受的待积当量剂量及其所致的待积有效剂量。用１９９４年呼吸道模型计算的结果低于１９７９年新肺模型的所得结果。     

Lumped pseudo fission products during burnup step in MCNP5-ORIGEN coupling system

Depletion calculation and accurate inventory of fission products in a nuclear system are required for criticality, safety and spent fuel management. Actual trend is to use Monte Carlo methods. It is well known that the fission process produces a large number of nuclides, some of which have a significant impact on the nuclear properties of the core and its behavior. In this study, we propose to determine the influence of fission products on the behavior of the IAEA 10 MW benchmark reactor. Even if nowadays we have powerful computing capability and we can solve the full system of fission products, such calculations are cumbersome and not needed because most of fission products have low absorption rates and therefore their precise concentrations calculation are not required. The practice is to identify and use only the nuclides which can have a significant absorption cross section.From the entire fission products of the available fissionable actinides, 214 nuclides have been considered. Their selection was essentially based on their absorption rates. To carry out the calculation, 81 were treated explicitly and 133 were lumped into pseudo fission products.A computational method has been developed for burnup and criticality calculations using MCNP5-ORIGEN coupling scheme. The MIXE_ACE program was developed and incorporated within this coupling scheme in order to mix and rewrite in ACE format the selected cross sections of the pseudo fission products for each burnup step. The mass weight of the constituent nuclides was used. The initial one group cross sections library for ORIGEN was generated using average flux spectrum in the core.Using the above methodology, an estimation of k_eff and cross sections during depletion calculations has been carried out for the IAEA 10 MW reactor based on UZrH1.6 fuel. The results are compared to those of ANL (Argonne National Laboratory), MCNP6 and other calculations by using selected fission products from WIMS library. Generally, the results are satisfactory but some discrepancies exist. The differences can be explained mainly by the nature of the fission products considered in the calculation and especially their cross sections.