Heats of Vaporization of UO2 and UO from UO2

In an earlier paper, a stochastic model of a power reactor has been proposed by the present author on the premise that the coolant-flow through a core is usually accompanied by random variations in the flow-rate, which are eventually largely responsible for the internal reactivity fluctuations.

In the present work, this model is extended to three different reactor systems: (a) where there exists a relaxation process corresponding to the effect of buoyant flow; (b) where a control or fuel element vibrates randomly, due to coolant flow-rate fluctuations; (c) where there are fluctuations in the inlet temperature with a non-white spectrum.

The noise spectra are derived for various state quantities with use made of the Langevin procedure. The theory is illustrated by referring chiefly to the neutron noise spectra, and comparing with the results of observations.

It is shown that the noise sources in question contribute significantly to the spectra, as compared with a low frequency component due to an inherent noise source in the coolant flow. In particular, a strong resonance peak of the spectra arises from the coupling between the random mechanical vibrations and coolant the flow-rate fluctuations.     

Chemical State of Fission Tellurium Released from Irradiated UO2 and UO2-Graphite Mixtures

The chemical state of fission Te released from ceramic fuels was studied by means of the thermochromatograph. In inert atmosphere, the fission Te was found to be released in the form of metal, but in the presence of oxygen, it readily reacted therewith to form TeO₂. The chemical state of I released from Te and TeO₂ was also determined.     

UO22+在羟基磷灰石上的吸附

以羟基磷灰石（HA）作骨体外模型结合计算机模拟的方法,研究了UO₂²⁺在HA上的吸附行为以及不同种类的螯合剂对其吸附行为的影响。结果表明,15 min后HA对UO₂²⁺的吸附达到平衡,HA颗粒形貌未发生明显改变,吸附过程符合拟二级动力学方程;乙二胺四乙酸（EDTA）和二乙基三胺五乙酸（DTPA）对HA上UO₂²⁺的促排作用类似,能与UO₂²⁺螯合形成大量的络合离子从而达到促排的目的,但会对骨骼无机相表面造成损伤;柠檬酸对吸附在HA上的UO₂²⁺具有较强的解吸作用,但高浓度的柠檬酸会造成HA颗粒形貌显著变化,因此不适合作为铀的促排剂;抗坏血酸对UO₂²⁺在HA上的解吸没有作用,而氨基三乙酸（NTA）会导致HA对铀的吸附率增加。     

Preparing UO2 kernels by gelcasting

A process named gel-casting has been developed for the production of dense UO2 kernels for the high-ten-temperature gas-cooled reactor. Compared with the sol-gel process, the green microspheres can be got by dispersing the U3O8 slurry in gelcasting process, which means that gelcasting is a more facilitative process with less waste in fabricating UO2 kernels. The heat treatment.     

Recoil Release of Fission-Gas from UO2

The fission-gas released from in-core UO₂ pellets was investigated with the helium-swept fuel irradiation facility installed in the Hitachi Training Reactor. It was found that the release rate of fission-gas increased gradually as irradiation continued for 5hr under a constant thermal neutron flux of 1.0×10¹² n/cm²·sec and at a fuel temperature of about 150°C. This observation indicates that a recoil-activated release mechanism may be more effective than direct-recoil release in the case of clad fuel in the region of temperature where fission-gas release rate is temperature independent.     

Thermal Conductivity of UO2-BeO Pellet

Beryllium oxide(BeO)-doped (0.3, 0.6, 0.9, 1.2 and 13.6 wt%) UO₂ pellets were fabricated to evaluate the effects of BeO precipitate shape on thermal conductivity. Precipitate distributions were of two types: BeO precipitated almost continuously along a grain boundary (designated BeO continuous type) and spherical BeO randomly dispersed within the matrix (designated BeO dispersed type). Thermal diffusivity was measured by a laser flash method and thermal conductivity was evaluated. The thermal conductivity increased with the BeO content. The thermal conductivity of the BeO continuous type was higher than that of the BeO dispersed type at lower temperatures, while their difference became smaller at higher temperatures. The thermal conductivities of UO₂-1.2 wt% BeO at 1,100K were higher than that of UO₂ by about 25 % for the BeO continuous type and by about 10 % for the BeO dispersed type. The thermal conductivities of both types were expressed by a semi-empirical equation as a function of volume fraction and shape of the BeO precipitates.     

Effects of UO2/Zircaloy-4 Mole Ratios on Reaction of UO2 Fuel with Molten Zircaloy-4

Dissolution of UO₂ crucibles by molten Zircaloy-4 (Zry) was investigated in the temperature range of 2,223-2,373 K and for specimens having UO₂/Zry mole ratios between 7 and 18.2. The uranium concentration in the Zry melt rapidly increased during initial reaction time and approached saturated values, depending on reaction temperature and UO₂/Zry mole ratio. Kinetics of uranium concentration increase in the melt was analyzed based on a natural convection mass transfer model that takes into account the change of contact surface area/melt volume ratio with reaction time. The saturated uranium concentration in the Zry melt was inversely proportional to the U0₂/Zry mole ratio. An empirical correlation of saturated uranium concentration in the Zry melt was obtained as a function of UO₂/Zry mole ratios and reaction temperature. This study of the empirical correlation was intended to estimate maximum UO₂ fuel dissolution by molten Zry cladding during severe fuel damage accidents for three different reactor type fuels.     

Distribution of Fission Products in Irradiated UO2

Inhomogeneities of fission products and plutonium distribution in irradiated UO₂ were examined. UO₂ pellets were ground so as to let them fracture along the grain boundaries, and the fragments were then dissolved successively into several fractions. The specific activities of these fractions were compared, and the distribution of fission products in the grain and grain boundaries were studied. It was found that at lower temperatures of irradiation (～1,400°C), the fission products, except Cs, were distributed fairly uniformly over the grain, while Cs had accumulated at the grain boundaries. At higher temperatures (near or above the melting point of UO₂), inhomogeneity was noticed in the Zr-Nb and Cs distribution. The concentration of these elements varied along the temperature gradient of the fuel rod than within the grains. Zr-Nb was found to be concentrated in the higher temperature areas of the fuel rod, while conversely, the concentration of Cs was smaller in the same areas. Distribution of Ce, Ru and Pu was found to be fairly uniform throughout the grains. The process of fission product migration in UO₂ is also discussed in the paper.     

Transient Thermal Characteristics of UO2 Fuel Rods

Transient characteristics of average temperature and heat flux of UO² fuel rods have been obtained with account taken of the effect of temperature dependence of thermal properties. The first order deviation of the temperature distribution inside the fuel rod is solved by modal expansion, and the coefficients and time constants obtained for the transfer functions. The results have been compared with those by the conventional method: the differences found as a result in the transfer functions for typical fuel rods are within 1 dB in gain and 5 degrees in phase when the volume-averaged fuel temperature is used in the conventional method.     

UO2中铌行为模拟研究

通过拟合Nb2O5的晶体结构,建立了铌的经验势.模拟计算孤立铌杂质表明:最近邻Nb-O间距为2.13A,均匀向铌靠近.近邻Nb-U间距为3.84(A),次近邻Nb-U间距为5.48(A),小于正常晶格的U-O间距以及U-U间距.铌附近间隙原子形成能全部增加.铌对周围氧原子施加了额外的束缚,氧间隙形成能随着Nb-O距离的增加而减小,但都大于完整的UO2的氧问隙形成能.引入少量铌元素后其影响范围扩大很多,采用静态过渡态理论计算的铌的空位机制扩散激活能为6.76eV,UO2中铌的扩散几乎不可能.多个铌杂质计算表明:铌离子倾向于团聚,形成替位缺陷簇.根据以上模拟计算,提出如下机制:铌以置换固溶体形式存在于UO2中,为了保持电中性,含铌的UO2中存在相应的氧间隙原子,氧化层总体上保持CaF2结构.铌形成二聚体或者四聚体,联结成网络,使氧的扩散势垒增大,降低了氧化速率.