The use of U3Si2 dispersed fuel in Low-Power Research Reactors

The use of U₃Si₂ as a Low Enriched Uranium (LEU) dispersed fuel in Low-Power Research Reactors is investigated in this paper. The fuel proves to be usable if some of the original fuel rods (HEU UAl₄–Al fuel) are still simultaneously employed (mixed core) without changing the structure of the actual core. About 3.5712 mk Initial Excess Reactivity (IER) is procured. Although the worths of both the control rod and the reactivity devices decrease, the safety of these reactors is higher in the case of the new LEU fuel. If the dimensions of the meat and/or the clad are allowed to change these reactors can be run with a meat 2.15 mm outer radius, and a clad 0.58 mm thickness. The IER will then be 4.1537 mk, and both the control rod (CR) worth and the safety margins decrease.     

Thermal compatibility studies of U3Si2 dispersion fuels prepared with centrifugally atomized powder

The interaction between atomized U₃Si₂ and aluminum in dispersion fuel samples has been characterized and compared with that of comminuted U₃Si₂. Fuel samples with atomized powder showed a smaller volume increase compared to those with the comminuted powder, irrespective of heat treatment, and volume fraction of U₃Si₂ powder. The possible reasons for this seem to be as follows: (1) the smaller specific surface area of the atomized spherical powder compared to the irregular comminuted powder translating in a smaller U₃Si₂–Al interface area for the former affecting what appears to be a diffusion-controlled interaction process, (2) the atomized fuel samples also contain lower fraction of as-fabricated porosity than the comminuted fuel samples, which may enhance the restraint force in the swelling fuel meat, (3) the comminuted powder particles have distinctive aluminum penetration paths in the form of deformation zones that originated from the comminution process. There appear to be two pronounced penetration paths of aluminum into atomized U₃Si₂ powder; (1) through the phase interface, leaving a central unreacted island, (2) along grain boundaries, leaving several unreacted islands.     

Post-irradiation examination of AlFeNi cladded U3Si2 fuel plates irradiated under severe conditions

Three full size AlFeNi cladded U₃Si₂ fuel plates were irradiated in the BR2 reactor of the Belgian Nuclear Research Centre (SCK·CEN) under relatively severe, but well defined conditions. The irradiation was part of the qualification campaign for the fuel to be used in the future Jules Horowitz reactor in Cadarache, France. After the irradiation, the fuel plates were submitted to an extensive post-irradiation campaign in the hot cell laboratory of SCK·CEN. The PIE shows that the fuel plates withstood the irradiation successfully, as no detrimental defects have been found. At the cladding surface, a multilayered corrosion oxide film has formed. The U-Al-Si layer resulting from the interaction between the U₃Si₂ fuel and the Al matrix, has been quantified as U(Al,Si)_4.6. It is found that the composition of the fuel particles is not homogenous; zones of USi and U₃Si₂ are observed and measured. The fission gas-related bubbles generated in both phases show a different morphology. In the USi fuel, the bubbles are small and numerous while in U₃Si₂ the bubbles are larger but there are fewer.     

U3Si2—Al弥散型燃料的辐照肿胀研究

介绍了U3Si2-Al弥散型燃料的辐照肿胀机理。将弥散型燃料的芯体视为连续基体中的微型燃料元件,应用裂变气体的行为机理描述燃料相中的气泡形成过程。研究结果表明：燃料相的肿胀引起燃料颗粒和金属基体之间的力学相互作用,金属基体能抑制燃料颗粒的辐照肿胀。在一定辐照条件下,本模型对燃料元件辐照肿胀的预测值与测量值相符。     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

新型U3Si2-Al燃料组件的工程验证试验研究

新研制的U₃Si₂-Al板状弥散型燃料组件结构复杂,国内对该燃料组件的结构材料、制造工艺、力学性能、运行特性等均缺少经验及评定标准。为得到该新型燃料组件的各种性能参数,开展了燃料包壳及结构材料的力学性能试验、燃料板及包壳材料的热物性及热稳定性试验、燃料板的力学性能试验、燃料板的正电子湮灭寿命试验、燃料组件的水力冲刷和解体试验等一系列的工程验证试验和专项研究,得到的各项实验数据为燃料组件的结构设计、可靠性分析、安全审评提供了重要依据,也为燃料组件的加工制造、堆内使用管理提供了借鉴。     

Formation of U2C3 in the reaction of UC2 with UO2

The formation of U₂C₃ by the reaction of UC₂ with UO₂ has been studied by chemical and X-ray analyses at temperatures between 1400 and 1700 °C in vacuo. The reaction is represented by 7 UC₂ + UO₂ → 4 U₂C₃ + 2 CO.     

含TBP-HNO3超临界CO2流体静态络合萃取U3O8

建立了一种快速降低萃取系统压力的静态络合萃取实验装置。在此装置上研究了含TBP-HNO₃超临界CO₂静态络合U₃O₈的快速气化测量方法,探索了含TBP-HNO₃超临界CO₂静态络合萃取U₃O₈的行为规律。     

A generalized model for radiation-induced amorphization and crystallization of U3Si and U3Si2 and recrystallization of UO2

A rate-theory model of radiation-induced amorphization and crystallization of U₃Si during ion irradiation has been generalized to include U₃Si₂ and UO₂. The generalized model has been applied to ion-irradiation and in-reactor experiments on U₃Si and U₃Si₂ and provides an interpretation for the amorphization curve (dose required to amorphize the material as a function of temperature), for the ion-radiation-induced nanoscale polycrystallization of these materials at temperatures above the critical temperature for amorphization, as well as for the role of the small crystallites in retarding amorphization. An alternative mechanism for the evolution of recrystallization nuclei is described for a model of irradiation-induced recrystallization of UO₂ wherein the stored energy in the UO₂ is concentrated in a network of sinklike nuclei that diminish with dose due to interaction with radiation-produced defects. The sinklike nuclei are identified as cellular dislocation structures that evolve relatively early in the irradiation period. The complicated kinetics involved in the formation of a cellular dislocation network are approximated by the formation and growth of subgrains due to the interaction of shock waves produced by fission-induced damage to the UO₂.     

An investigation on the green properties of U-10wt.%Mo/Al and U3Si2/Al powder compacts

The effects of fuel powder volume fraction and fuel particle shape on green properties of compacts, which were produced by processing the blended U-10wt.%Mo and U₃Si₂ with Al powders were investigated respectively, with respective to the compacting pressure range of 50–400 MPa. The relative density of the compacts increases with decreasing volume fraction of fuel powder. The compressibility of comminuted powder compacts was larger than that of the atomized powder compacts due to the fragmentation of comminuted particles, and the compressibility of the compacts of U-10wt.%Mo was larger than that of the compacts of U₃Si₂ due to the deformation of U-10wt.%Mo particles. The green strength of the comminuted powder compacts is higher than that of the atomized powder compact. This seems to have resulted from the smaller pore size and the larger contact area between the comminuted fuel powders and Al powders. It is suggested that the compacting condition adjustment be required to fabricate the atomized powder compacts having comparable green strength.     

Fission product release in high-burn-up UO2 oxidized to U3O8

Results of oxidation experiments on high-burn-up UO₂ are presented where fission-product vaporisation and release rates have been measured by on-line mass spectrometry as a function of time/temperature during thermal annealing treatments in a Knudsen cell under controlled oxygen atmosphere. Fractional release curves of fission gas and other less volatile fission products in the temperature range 800-2000 K were obtained from BWR fuel samples of 65 GWd t⁻¹ burn-up and oxidized to U₃O₈ at low temperature. The diffusion enthalpy of gaseous fission products and helium in different structures of U₃O₈ was determined.     

Cuboctahedral oxygen clusters in U3O7

When UO₂ is oxidised to U₃O₇, the positions in the crystal lattice of all the uranium atoms and of about 70% of the oxygen atoms are hardly affected. The remaining oxygen atoms occupy new sites which are located 310 pm along 〈1 1 0〉 vectors from the holes in the fluorite framework of UO₂. These results, which are based on the analysis of neutron diffraction powder data, are consistent with the concept that excess oxygen in U₃O₇ is accommodated in cuboctahedral anionic clusters.     

Thermoelectric power studies of ferromagnet U2ScB6C3

The thermoelectric power (TEP) of a ferromagnet U₂ScB₆C₃ (T_C = 61 K) has been measured in the temperature range 5-300 K. The TEP is positive over the whole measured temperature range and reaches a relatively large value at room temperature of 29 μV/K. Below 30 K and above 200 K the TEP follows a straight line S(T) ∼AT, with slope of 0.23 and 0.085 μV/K², respectively. The change in the slope can be explained by the electron-phonon interaction renormalization effects or spin-reorientation associated with a change in the electronic structure. Analysing the temperature dependence of the ratio [S(T)/T]/[S_300 K/300] and taking into account the specific heat data, we suggest that spin fluctuations are another important factor in determining the thermoelectric power behaviour of U₂ScB₆C₃.     

Test irradiations of full-sized U3Si2-Al fuel plates up to very high fission densities

In the course of the licensing procedure of the ‘Forschungsneutronenquelle Heinz Maier-Leibnitz’, i.e. the new 20 MW high-flux research reactor FRM II in Garching near Munich, extensive test irradiations have been performed to qualify the U₃Si₂-Al dispersion fuel with a relatively high density of highly enriched uranium (93 wt% of ²³⁵U) up to very high fission densities. Two of the three FRM II type fuel plates used in the irradiation tests contained U₃Si₂-Al dispersion fuel with HEU densities of 3.0 gU/cm³ or 1.5 gU/cm³ (‘homogeneous plates’) and one plate had two adjacent zones of either density (‘mixed plate’). They were irradiated in the French MTR reactors SILOE and OSIRIS in the years before 2002. The local plate thickness was measured on certain tracks along the plates during interruptions of the irradiation. The maximum fission density obtained in the U₃Si₂ fuel particles was 1.4 × 10²² f/cm³ and 1.1 × 10²² f/cm³ in the 1.5 gU/cm³ and 3.0 gU/cm³ fuel zones, respectively. In the course of the irradiations, the plate thickness increased monotonically and approximately linearly, leading to a maximum plate thickness swelling of 14% and 21% and a corresponding volume increase of the fuel particles of 106% and 81%, respectively. Our results are discussed and compared with the data from the literature.     

Characterization and densification studies on ThO2-UO2 pellets derived from ThO2 and U3O8 powders

ThO₂ containing around 2-3% ²³³UO₂ is the proposed fuel for the forthcoming Indian Advanced Heavy Water Reactor (AHWR). This fuel is prepared by powder metallurgy technique using ThO₂ and U₃O₈ powders as the starting material. The densification behaviour of the fuel was evaluated using a high temperature dilatometer in four different atmospheres Ar, Ar-8%H₂, CO₂ and air. Air was found to be the best medium for sintering among them. For Ar and Ar-8%H₂ atmospheres, the former gave a slightly higher densification. Thermogravimetric studies carried out on ThO₂-2%U₃O₈ granules in air showed a continuous decrease in weight up to 1500 °C. The effectiveness of U₃O₈ in enhancing the sintering of ThO₂ has been established.     

The mechanism and kinetics of U2C3 formation by the synthetic reaction

In fast breeder reactors it is planned to use the fuel in the form of (U, Pu)C with a slight carbon hyperstoichiometry. It is therefore important to know under what conditions the synthetic reaction $\text{UC + UC}{}_2\text{U}{}_2\text{C}{}_3$ occurs, since the hyperstoichiometric carbon, which exists as a uranium dicarbide phase (UC₂) after manufacture, is to be converted to U₂C₃. The literature concerning the reaction conditions is partly contradictory. In this paper, the influence of thermal or mechanical pretreatments on the formation of U₂C₃ was investigated experimentally and is discussed in connection with other published data. It was found that the relative increase of the U₂C₃ nuclei density caused by grinding corresponds to the increase in surface of the ground material. A quantitative kinetic examination of the U₂C₃ formation was made and the activation energy for the synthetic reaction in powder was found to be $\text{394 ± 30}$ kJ/mol.     

The kinetics of U3Si formation in cast U3Si alloy

The growth of U₃Si in cast U-3.8 wt % Si alloy was measured by determining both the U₃Si₂/U₃Si interface movement, and the change in the total amount of the various phases. The activation energy controlling growth was found to be 50 ± 1 kcal/mole.     

A dilatometric study of the solubility of U4O9 in UO2

Phase relationships in the system UO₂-O₉ were studied using a dilatometei in which the O/U ratio of a UU_2+x specimen could be both controlled and measured. Phase boundary temperatures were indicated by changes in expansion or contraction rate during heating or cooling, respectively. The solubility of U₄O₉ in UO₂ agreed with the results of previous workers using other techniques. The dependence of solubility on temperature is complex, and appears to be influenced by a high-temperature phase transition in U₄0₉.     

XRD investigation of ion irradiated Ti3Si0.90Al0.10C2

Ti₃SiC₂ is one of the most promising materials belonging to M_n+1AX_n phases, which exhibit good damage tolerance, thermal stability and mechanical properties.Recently, in the frame of research on future gas cooled fast nuclear reactors, Ti₃SiC₂ has been considered as an innovative candidate material, which could be incorporated in some core components such as fuel cladding. At the present time, however, very few data are available concerning the behaviour of this material after irradiation. In this work, Ti₃Si_0.90Al_0.10C₂ samples were irradiated with high energy Kr and Xe ions and characterized by X-ray diffraction. Patterns were analysed in terms of change in peak intensity, peak position and width. Rietveld refinements were also performed. Increase in micro-strains and lattice parameter with irradiation dose was highlighted. The formation of β-Ti₃SiC₂, which has never been observed by experimental XRD on non irradiated material, was proposed for the highly irradiated samples. A partial recovery of the microstructure with temperature was found.