Fission Gas Release from Uranium Heated in Carbon Dioxide

An experiment to predict the amount of fission gas release from U heated in CO₂ was made in connection with the safely evaluation of the Tokai Atomic Power Reactor. Fission gas release from slightly irradiated U pieces was measured as a function of percent oxidation, and it was found that the percent release was proportional to the percent oxidation. The fission gas released before the U was completely oxidized was 11%, 23% and 48% at 500°, 600° and 700°C, respectively. The form of temperature dependence of the fission gas release was of the Arrhenius type, with an activation energy of 5.1kcal/mol.     

Oxidation of Uranium by Partially Decomposed Carbon Dioxide

The oxidation rate of U rods by partially decomposed CO₂ was measured thermogravimetrically and compared with that by undecomposed CO₂, in an attempt to study the effect of ionizing radiation on the compatibility of U with CO₂. The decomposition was by silent discharge. Measurement was performed for cases of 0.29%, 0.44% and 0.68% decomposed CO₂, at 500° and 600°C. It was found that the oxidation rate rises almost linearly with increasing percentage of decomposition of the CO₂. At 500°C the oxidation rate with 0.68% decomposed CO₂ was eighteen times higher than with undecomposed CO₂, and five times higher at 600°C. Such increase was related to O₂ content in the decomposed gas.     

烧结二氧化铀燃料裂变气体肿胀的数值模拟研究

建立低温条件下烧结二氧化铀燃料(简称UO2燃料)中裂变气体的肿胀计算模型,采用有限差分方法编写计算程序,定量计算不同燃耗和温度条件下UO2燃料中固溶态的裂变气体份额、裂变气体气泡的密度与平均半径以及它们对燃料肿胀的贡献.计算表明,该模型能用于预测低温条件下UO2燃料中裂变气体所导致的肿胀随燃耗的变化规律.     

气体扩散法分离二氧化碳实验研究

随着碳13呼气检测实验的推广和普及,碳13同位素产品的市场需求呈上升趋势.为探索碳13同位素的工业化分离方法,开展以二氧化碳为介质分离碳同位素的气体扩散分离实验研究.根据二氧化碳的平均自由程,使用平均孔径55～65 nm的有机高分子多孔膜作为分离膜,并开发了单级分离装置.通过单级扩散分离实验,探究分离膜层数和膜前压强对...     

Oxidation of Uranium Dioxide

The oxidation of UO₂ was studied by thermogravimetry and X-ray diffraction. It was clarified that the thermal history covering the first stage of the oxidation from UO₂ to U₃O UO₇ significantly influenced the rate of the oxidation of the second stage from U₃O₇ to U₃ O₈.

The entire oxidation reaction proceeded in what to all appearances, was a single stage when the specimen temperature was raised rapidly, whereas at slower rates of heating up, two distinct stages of oxidation were observed, separated by an intermediate induction period. These findings suggest the existence of a close connection between the rate of formation of the U₃O₇ phase and the rate of the subsequent oxidation of this phase: A slow formation of U₃O₇ would tend to prolong the induction period preceding the second stage of the oxidation. A similar effect was observed also with annealing of the intermediate U₃O₇ at 200°C: The increase of annealing time prolonged the induction stage.

The rate of the second stage oxidation was fairly well expressed by Johnson & Mehl's equation, log (1/(1-y/)=(1/2.303)kⁿtⁿ . The time exponent n in this equation varied in the range of 1.0~2.5, and the rate constant k of 1.15×10^?4~2.04 ×10^?1 min^?1, depending on the experimental conditions.     

UO2粉末低温氧化表面改性研究

本工作研究探索UO2粉末的低温预氧化工艺.分别以10和20 ℃/min的升温速度,将UO2粉末于流动空气中加热至900 ℃进行热分析,试验观测到第1个放热峰温度为224.6和239.7 ℃,第2个放热峰温度为367.0和381.7 ℃.采用X射线衍射法分析了UO2粉末于240、382和815 ℃下氧化保温8 h后超化学计量铀氧化物UO2 x的物相组成.结果表明:UO2粉末在100～120 ℃下失去吸附水;在240 ℃氧化8 h后,有极少量的U3O7生成,382和815 ℃下氧化8 h后有U3O8生成.欲实现UO2低温烧结,获得2.25的最佳O/U原子比,可将UO2粉末预氧化工艺温度控制在240～370 ℃之间.     

二氧化铀粉末513K下空气氧化机制和动力学研究

为获取壳层O/U比约为2.25而芯部仍为UO₂的预氧化活性粉末以进行UO₂芯块的低温烧结试验,研究了UO₂粉末在513 K下于静态空气中的氧化机制和动力学。结果表明：UO₂粉末在513 K下氧化,可分为氧化前期和氧化后期;UO₂粉末氧化后期绝对增重比率极限为0.489%;氧化前期反应速率受界面化学反应控制,氧化后期反应速率受扩散控制;氧化前期,UO₂粉末氧化属于缩核模型,UO₂单颗粒单位面积上氧化质量的增加量解析式为Dw/Sn＝0.104 1t－2.146 67,氧化速率常数k_c＝3.49×10⁶exp(－73 919.39/RT)。     

以二氧化碳为介质扩散分离碳同位素的可行性研究

为进一步提高气体扩散法分离¹³C同位素的效率,在前期初步实验的基础上,开展单级扩散分离参数优化实验研究,并进行高丰度¹³C同位素制备的级联方案初步设计。在相对优化的实验参数条件下,气体扩散分离二氧化碳的基本全分离系数可以达到1.01以上。对单级分离实验数据进行计算,初步拟合出供料流量与膜前后压强的函数关系。采用多元分离理论对扩散分离二氧化碳进行级联分析计算,以天然二氧化碳为原料,可通过两次级联分离获得高丰度¹³C同位素。第一次阶梯级联分离的重馏分¹³C同位素丰度大于42%,并将其作为第二次阶梯级联分离的供料,第二次阶梯级联分离的轻馏分¹³C同位素丰度大于90%。     

Separation of Fission Products from Irradiated Uranium Dioxide in Fused Nitrate Systems

The separation of fission products from irradiated UO₂ in fused nitrate systems was studied by the following procedure:

Dissolution of UO₂ in fused NH₄NO₃; adsorption of fission products on glass powder; fused salt chromatography with γ-Al₂O₃; evaporation from fused nitrates; and precipitation of U-compound (probably alkali uranate, M₂U₂O₇) in a fused LiNO₃-KNO₃ mixture. Radiochemically pure Zr-Nb was selectively separated from the fused NH₄NO₃ melt by adsorption. Ru and I were completely evaporated from fused LiNO₃-KNO₃ mixture at above 280°C and from fused NH₄NO₃at above 240°C, respectively. By means of chromatography with a γ-Al₂O₃ column, Zr-Nb and rare earths were completely separated from U in fused NH₄NO₃ or fused NH₄NO₃-LiNO₃ medium at 150° to 180°C, and further, Cs, Ba and Ru were also expected to be separable from U under suitable condition. When the U-compound was precipitated in fused LiNO₃-KNO₃ at 350°C, Cs, Sr, Ba and Ru were fairly well separated from U.     

Fission Product Gas Release in Burst Region of Uranium Monocarbides

Uranium enrichment in a thermally driven countercurrent gas centrifuge is studied theoretically on the basis of dander's separation theory and the short-bowl flow solution proposed recently by Sakurai and Matsuda. The separation factor is computed and compared with experimental results by Groth. Agreement of the computed results with experiment is fairly good, except for very low feed rates. Further improvements are discussed.     

超临界CO2清洗铀屑技术

利用超临界CO2对铀机加工后切屑上残留的冷却液进行清洗,并通过称重结合化学分析法对清洗效果进行评价.从初步实验结果可知,提高清洗温度、压力,延长清洗时间,加大流速,都能够在一定程度上提高对切屑上冷却液的清洗效果.在超临界CO2中引入超声波,可进一步提高清洗效果.最后,对清洗工艺参数进行了初步优化.     

Carbon Diffusion in Uranium during Thermal Reprocessing in Vacuum

It is shown that during annealing of uranium samples at 550–950°C in a 10^–3 Pa vacuum carbon impurity diffuses from the interior volume to the surface of the samples. The diffusion flux of carbon directed toward the surface arises as a result of the formation of uranium oxymonocarbide, which contains carbon impurity, on the surface in the presence of low oxygen partial pressure. The carbon diffusion rate in -uranium grains depends on the crystallographic orientation of the grains, i.e., carbon diffusion in -uranium is anisotropic.     

超声波辅助超临界CO_2处理铀表面技术

开展了超声波对超临界CO2清洗铀样品表面污物效率影响的研究,探讨了利用超声波改善铀样品抗大气腐蚀效果的可行性。研究结果表明,向超临界CO2介质中引入超声波后,超临界CO2对样品表面的机油、水和三乙醇胺的清洗能力和效率都得到了显著提高。用俄歇电子能谱和XRD对样品表层成分及物相的分析结果表明,经过含φ=0.5%羰基镍的超临界CO2结合超声波处理的铀样品,表层一定深度范围内有镍元素存在。热氧化增重实验证实,经过超声波作用的含羰基镍超临界CO2处理后,铀样品抗大气腐蚀能力比仅用超临界CO2处理有明显增强。     

燃料氧化对裂变产物扩散释放的影响研究

分析燃料氧化对裂变产物扩散释放的影响,得出结论:燃料氧化后,燃料中铀原子空位增多对裂变产物释放的影响要大于导热率降低后燃料温度上升带来的影响;裂变产物释放份额随氧铀比升高而增大,燃料棒线功率密度越高,效应越明显。     

超临界二氧化碳清洗铀样品技术研究

针对铀样品在机械加工过程中残留的几种常见冷却液成分：水、机油和三乙醇胺，开展了超临界二氧化碳对铀样品的清洗实验，探讨了超临界二氧化碳对铀样品表面常见机械加工残留物的清洗效果，同时对经超临界二氧化碳浸泡的铀样品表面化学成分及相结构也进行了分析。清洗实验结果表明，超临界二氧化碳流体对水和机油有一定的溶解能力，具有较好的清洗效果，但对三乙醇胺的溶解能力较差，几乎不溶解。对三乙醇胺可通过向超临界二氧化碳引入超声波的方法来有效清除。铀表面分析结果显示，清洗后的铀表面只有U和UO2相；俄歇电子能谱显示，铀与超临界二氧化碳作用过程中，碳和氧进入到了铀表层的一定深度范围。腐蚀实验结果表明，铀在超临界二氧化碳中浸泡后，其抗大气腐蚀的能力不会减弱，相反还有一定程度的提高。     

快中子诱发裂变测定铀同位素丰度

铀同位素丰度分析是核燃料循环中重要的分析项目。本工作在中子活化基础上提出用T(d,n)⁴He反应产生的快中子诱发裂变铀样品来测定铀同位素丰度的分析方法。该方法利用铀同位素丰度与特定裂变产物的平均产额比呈指数关系的特点,选择〖AKY-〗(⁹²Sr)/〖AKY-〗(¹⁰⁵Ru)、〖AKY-〗(⁹²Sr)/〖AKY-〗(¹³⁵I)与铀同位素丰度间的关系曲线作为工作曲线,并对模拟样品进行分析。结果表明,〖AKY-〗(⁹²Sr)/〖AKY-〗(¹⁰⁵Ru)、〖AKY-〗(⁹²Sr)/〖AKY-〗(¹³⁵I)计算结果与模拟样品的相对偏差分别为0.8%和1.8%,RSD优于3%,与γ能谱法的分析结果相符合。     

Fluorination of Uranium Dioxide by Bromine-Fluorine System

The addition of bromine to fluorine flow makes it possible to fluorinate UO₂ into UF₆even below 200°C, at which temperature fluorination does not proceed with fluorine alone. Presence of bromine corresponding to about 6% of the fluorine concentration is sufficient to induce the fluorination. This effect of bromine is much greater than would be expected from the presence of bromine fluorides in concentrations such as would be produced by direct combination of the fluorine with the added bromine. It appears that the fluorination is enhanced when the mixed gas is held for 3~20 sec before arriving at the sample. The main component of the reactant gas is BrF₃.     

Separation of Uranium from Fission Products

A simple procedure has been developed for separating U from F.P. and neutron-irradiated Th. The separation is performed with tri-n-butyl phosphate in a system of dodecane-mixture of sulfuric acid and aluminium nitrate.

Uranium dioxide was irradiated with 20 MeV bremsstrahlung, which produced both ²³⁷U and F.P. The target was dissolved in dilute nitric acid and U was extracted into the organic phase of the above mentioned system. Finally, U in the organic phase was back-extracted into an aqueous phase. The γ-ray spectrum and decay curve of the separated U fraction show no radioactive nuclides other than U isotopes and its decay products.

This method can also be applied to the preliminary separation of ²³⁸U from neutron- irradiated Th.

The distribution ratios (Kd) for U, Th and some other elements in the extraction system are also given.     

采用椭偏光谱技术研究铀氧化动力学

采用椭偏技术研究了45～95 ℃和5 kPa纯氧气氛中金属铀的氧化,通过不同温度下金属铀表面氧化层厚度随时间的变化规律,得到贫铀在初始氧化阶段的表面氧化层厚度与时间的关系均呈现抛物线规律;绘制出了在不同温度下金属铀表面的氧化动力学曲线,由此得到在低于100 ℃, 5 kPa的纯氧气氛下贫铀表面形成氧化物的反应活化能为87.165 kJ/mol.     

Range Calculation of Fission Fragments in Uranium Coating

<正>Fission chambers use ~(235)U as neutron sensitive material,and energy deposition of fissile isotope fragments produces the electrical signal in the following circuits.In order to calculate the self absorption effect of the fission fragments in the coating,fission fragments range calculation