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.     

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.     

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.     

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.     

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。     

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.     

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     

Microstructure of Irradiated Uranium Dioxide

Sintered UO₂ pellets irradiated at high temperature to a burnup of 2,700 MWD/T-UO₂ were studied by replica electron microscopy. The as-irradiated pellet surfaces were covered by a thin layer easily soluble in dilute acid. This layer apparently was deposited as a sublimate during irradiation. The etched surfaces of the irradiated specimens were uniformly dispersed with flat circular domes about 1 μ in diameter, constituted of precipitates resistant to etching solution, apparently a mixture of fission products and UO₂. On the surface of each dome were several “sprout” precipitates 0.1 to 0.3 μ in diameter, and almost immune to chemical etchant.     

以乙醇为介质气体扩散法分离碳同位素的可行性研究

碳-13作为同位素示踪技术的标记物,应用广泛,市场需求呈上升趋势。为探索碳-13同位素的分离方法,开展以乙醇为介质的气体扩散分离实验,并在单级实验的基础上进行级联计算。单级分离实验的结果表明,在现有的实验条件下,气体扩散法分离乙醇的基本全分离系数可达1.0089,以乙醇为介质扩散分离碳同位素可行。通过级联计算可知,以天然乙醇为原料,结合分离的可行性和经济性,经过一次矩形级联或相对丰度匹配级联分离,能够得到碳-13同位素丰度大于25%的重组分。如能将碳-13同位素丰度大于25%的乙醇转化为合适形态的碳化合物,可再进一步分离得到更高丰度的碳-13同位素。     

Preparation of Uranium Plating For Fission Ionization Chamber

<正>Using electro-deposition method to make uranium plating for fission ionization chamber is characterized by small volume solution,short operating time and less radioactive waste material.This paper mainly introduces the technical research of electrodes plating process.And the     

Fluorination of Uranium Metal by Fluorine Gas

A study has been made on the fluorination of uranium metal chips to UF₆ with fluorine gas in the temperature range of 100° to 400°C. The formation of UF₆ was influenced by the concentration and the flow rate of fluorine gas and by the reaction temperature. The reaction occurred apparently above 200°C. Uranium metal was first converted to low fluorine content compound such as UF₃ or UF_4-x, then to UF₄ and finally to UF₆. The intermediate compounds were confirmed by X-ray analysis and by their color.     

超临界CO2处理后的金属铀表面氧化腐蚀和电化学腐蚀实验研究

采用重量法和电化学方法分别得到表面经超临界CO2处理后的铀试样在60℃、70％RH条件下的氧化动力学曲线(△m-t)和在50μg/g Cl^-溶液中的阳极极化曲线、自腐蚀电流等抗蚀评价指标。结果表明：经超临界CO2处理后，金属铀表面抗氧化腐蚀性能和抗化学介质腐蚀性能均有一定程度提高。对钝化膜提高金属铀表面抗腐蚀性能的原因进行了探讨。     

Oxidation Behavior of Uranium Monosulfide

The oxidation behavior of uranium monosulfide powder in oxygen and air has been studied over a range from room temperature to 900°C, by means of DTA, TGA, X-ray diffraction and chemical analysis. Most of the reactions occurring during the oxidation process have been identified. Intermediate oxidation products observed are γ-US₂, UOS, UO₂SO₄, UO₂ and UO_2+x.     

Diffusion of Fission-Produced Xenon in Uranium Monocarbide

The diffusion of Xe atoms in UC at temperature from 800° to 1,400°C was studied on samples irradiated to various doses up to 2.9×10¹⁸ nvt. The apparent activation energy of the diffusion was found to vary with fluence, which corresponded to the extent of neutron or fission fragment damage in the specimen.

The diffusion was observed to be enhanced at exposures below 10¹⁷ nvt. At the higher doses, on the other hand, trapping of the gas atoms by defect clusters became evident, which inhibited the gas release.

When the irradiated samples were once heated up to 1,400°C, the activation energy of Xe diffusion through them showed a constant value of 83±5 kcal/mol, which was independent of the neutron fluence.

The scattered values of activation energy reported in literature are explained from the present results as resulting from differences in the activation energy of Xe diffusion, which depends on the extent of radiation damage retained in the UC.     

The Behavior of Uranium Dioxide in Various Gases

The behavior of uranium dioxide in an oxidative medium in air and a neutral medium in argon for several thousands of hours was studied. Negligible additional oxidation of pellets occurred at room temperature; this was indicated by an increase of the stoichiometric mixture ratio of uranium dioxide near the surface. The oxidation rate constant at 295 K was k = 2.77·10^–7 in air and 5·10^–8 g/(cm²·h) in argon. For oxidation in air at 295–503 K only solid solutions of oxygen in uranium dioxide were formed. The temperature dependence of the oxidation rate constant was determined. The threshold temperature above which higher oxides of uranium form was determined. The stoichiometric mixture ratio has no effect on the oxidation rate constant of uranium dioxide in air. 1 figure, 3 tables, 9 references.