The ternary system: silicon-uranium-vanadium

Phase equilibria in the system Si-U-V were established at 1100 °C by optical microscopy, EMPA and X-ray diffraction. Two ternary compounds were observed, U₂V₃Si₄ and (U_1−xV_x)₅Si₃, for which the crystal structures were elucidated by X-ray powder data refinement and found to be isotypic with the monoclinic U₂Mo₃Si₄-type (space group P2₁/c; a = 0.6821(3), b = 0.6820(4), c = 0.6735(3) nm, β = 109.77(1)°) and the tetragonal W₅Si₃-type (space group I4/mcm, a = 1.06825(2), c = 0.52764(2) nm), respectively. (U_1−xV_x)₅Si₃ appears at 1100 °C without any significant homogeneity region at x ∼ 0.2 resulting in a formula U₄VSi₃ which corresponds to a fully ordered atom arrangement. DTA experiments clearly show decomposition of this phase above 1206 °C revealing a two-phase region U₃Si₂ + V₃Si. At 1100 °C U₄VSi₃ is in equilibrium with V₃Si, V₅Si₃, U₃Si₂ and U(V). At 800 °C U₄VSi₃ forms one vertex of the tie-triangle to U₃Si and V₃Si. Due to the rather high thermodynamic stability of V₃Si and the corresponding tie-lines V₃Si + liquid at 1100 °C and V₃Si + U(V) below 925 °C, no compatibility exists between U₃Si or U₃Si₂ and vanadium metal.     

Reactivity temperature coefficients for the HEU and LEU fuel of the MNSR reactor

Analysis of the Reactivity Temperature Coefficients of the Miniature Neutron Source Reactor (MNSR) for normal and accidental conditions (above 45 °C) using HEU-UAl₄ and the LEU: U₃Si, U₃Si₂ and U9Mo fuel were carried out in this paper. The Fuel Temperature Coefficient (FTC), Moderator Temperature Coefficient (MTC), and Moderator Density Coefficient (MDC) were calculated using the GETERA code. The contribution of each isotope presented in the fuel cell was calculated for the temperature range of 20 °C–100 °C at the beginning of the core life. The average values of the FTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were found to be: −2.23E-03, −1.85E-02, −1.96E-02, −1.85E-02 mk/°C respectively. The average values of the MTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −8.91E-03, −1.24E-04, −4.70E-03, 2.10E-03 mk/°C respectively. Finally, the average values of the MDC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −2.06E-01, −2.03E-01, −2.04E-01, −2.03E-01 mk/°C respectively. It's found also that the dominant reactivity coefficient for all types of fuel is the MDC.     

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl_x–Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si–Al and U₃Si₂–Al) and oxide (U₃O₈–Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 °C to 50 °C and 100 °C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U₃O₈–Al was about 2% more than the original UAl_x–Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.     

Microstructural characterization of U-7Mo/Al-Si alloy matrix dispersion fuel plates fabricated at 500 °C

The starting microstructure of a dispersion fuel plate will impact the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of U-Mo dispersion fuel plates, particularly the interaction layers that can form between the fuel particles and the matrix, scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses have been performed on samples from depleted U-7Mo (U-7Mo) dispersion fuel plates with either Al-2 wt.% Si(Al-2Si) or AA4043 alloy matrix. It was observed that in the thick interaction layers, U(Al, Si)3 and U6Mo4Al43 were present, and in the thin interaction layers, (U, Mo) (Al, Si)3, U(Al, Si)4, U3Si3Al2, U3Si5, and possibly USi-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this investigation, the time that a dispersion fuel plate is exposed to a relatively high temperature during fabrication will impact the nature of the interaction layers around the fuel particles. Uniformly thin, Si-rich layers will develop around the U-7Mo particles for shorter exposure times, and thicker, Si-depleted layers will develop for the longer exposure times.     

Kr ion irradiation study of the depleted-uranium alloys

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si)₃, (U, Mo)(Al, Si)₃, UMo₂Al₂₀, U₆Mo₄Al₄₃ and UAl₄. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10¹⁹ ions/m² (∼10 dpa) with an Kr ion flux of 10¹⁶ ions/m²/s (∼4.0 × 10⁻³ dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.     

Heats of formation of (U,Mo)Al3 and U(Al,Si)3

The heats of formation of (U,Mo)Al₃ intermetallic compounds were obtained by measuring the reaction heats of U-Mo/Al dispersion samples by differential scanning calorimetry. Based on literature data for the reaction heats of U₃Si/Al and U₃Si₂/Al dispersion samples, the heats of formation of U(Al,Si)₃ as a function of the Si content were calculated. The heat of formation of (U,Mo)Al₃ becomes less negative as the Mo content increases. Conversely, the heat of formation of U(Al,Si)₃ becomes more negative with increasing Si content.     

Post-irradiation examination of uranium-7 wt% molybdenum atomized dispersion fuel

Two low-enriched uranium fuel plates consisting of U-7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK•CEN. The plates were submitted to a heat flux of maximum 353 W/cm² while the surface cladding temperature is kept below 130 °C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% ²³⁵U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al₃ and (U,Mo)Al₄. Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat à l’énergie atomique (CEA).     

Temperature and dose dependence of fission-gas-bubble swelling in U3Si2

Large fission gas bubbles were observed during metallographic examination of an irradiated U₃Si₂ dispersion fuel plate (U0R040) in the Advanced Test Reactor (ATR). The fuel temperature of this plate was higher than for most of the previous silicide-fuel tests where much smaller bubble growth was observed. The apparent conditions for the large bubble growth are high fission density (6.1 × 10²¹ f/cm³) and high fuel temperature (life-average 160 °C). After analysis of PIE results of U0R040 and previous ANL test plates, a modification to the existing athermal bubble growth model appears to be necessary for high temperature application (above 130 °C). A detailed analysis was performed using a model for the irradiation-induced viscosity of binary alloys to explain the effect of the increased fuel temperature. Threshold curves are proposed in terms of fuel temperature and fission density above which formation and interconnection of bubbles larger than 5 μ are possible.     

Low silicon U(Al,Si)3 stabilization by Zr addition

Previous knowledge states that (U,Zr)Al₃ and U(Al,Si)₃ phases with Zr and Si content higher than 6 at.% (7.7 wt%) and 4 at.% (1.4 wt%), respectively, does not partially transform to UAl₄ at 600 °C. In this work, four alloys within the quaternary system U-Al-Si-Zr were made with a fixed nominal 0.18 at.% (0.1 wt%) Si content in order to assess the synergetic effect of both Zr and Si alloying elements to the thermodynamic stability of the (U,Zr)(Al,Si)₃ phase. Heat treatments at 600 °C were undertaken and samples were analyzed by means of XRD, EPMA and EDS techniques. A remarkable conclusion is that addition of 0.3 at.% Si in the (U,Zr)(Al,Si)₃ phase reduces in 2.7 at.% the necessary Zr content to inhibit its transformation to U(Al,Si)₄.     

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.     

Characterization of the reaction layer in U-7wt%Mo/Al diffusion couples

The reaction layer in chemical diffusion couples U-7wt%Mo/Al was investigated using optical and scanning electron microscopy, electron probe microanalysis and X-ray diffraction (XRD) techniques. When the U-7wt%Mo alloy was previously homogenized and the γ(U, Mo) phase was retained, the formation of (U, Mo)Al₃ and (U, Mo)Al₄ was observed at 580 °C. Also a very thin band was detected close to the Al side, the structure of the ternary compound Al₂₀UMo₂ might be assigned to it. When the decomposition of the γ(U, Mo) took place, a drastic change in the diffusion behavior was observed. In this case, XRD indicated the presence of phases with the structures of (U, Mo)Al₃, Al₄₃U₆Mo₄, γ(U, Mo) and α(U) in the reaction layer.     

Transmission electron microscopy investigation of irradiated U-7 wt%Mo dispersion fuel

The microstructural evolution of atomised U-7 wt%Mo alloy fuel under irradiation was investigated by transmission electron microscopy on material from the experimental fuel plates used in the FUTURE irradiation. The interaction layer that forms between the U(Mo) particles and the Al matrix is assumed to become amorphous under irradiation and as such cannot retain the fission gas in stable bubbles. As a consequence, gas filled voids are generated between the interaction layer and the matrix, causing the fuel plate to pillow and finally fail. The present analysis confirms the assumption that the U(Mo)-Al interaction layer is completely amorphous after irradiation. The Al matrix and the individual U(Mo) particles, with their cellular substructure, have retained their crystallinity. It was furthermore observed that the fission gas generated in the U(Mo) particles has formed a bubble superlattice, which is coherent with the U(Mo) lattice. Bubbles of roughly 1-2 nm size have formed a 3-dimensional lattice with a lattice spacing of 6-7 nm.     

Thermal–hydraulics features of the use of LEU U3Si2 fuel in low-power reactors

Calculations for the use of the U₃Si₂ LEU fuel in low-power research reactors were made. The design basis accident was simulated using the feedback coefficients calculated by the BMAC system. Usability of this fuel in low-power research reactors was demonstrated for both normal daily and accidental operation conditions even if the power of the reactor touches 142 kW during the design basis accident simulation. Both HEU and LEU fuels behave similarly in the normal operation, the temperature of the cladding reaching about 60 °C while higher temperature are obtained for the accidental conditions in the case of the LEU fuel (about 113.7 °C against 98.6 °C for the fuel center temperatures).     

Transmission electron microscopy characterization of irradiated U–7Mo/Al–2Si dispersion fuel

The plate-type dispersion fuels, with the atomized U(Mo) fuel particles dispersed in the Al or Al alloy matrix, are being developed for use in research and test reactors worldwide. It is found that the irradiation performance of a plate-type dispersion fuel depends on the radiation stability of the various phases in a fuel plate. Transmission electron microscopy was performed on a sample (peak fuel mid-plane temperature ～109 °C and fission density ～4.5 × 10²⁷ f m^?3) taken from an irradiated U–7Mo dispersion fuel plate with Al–2Si alloy matrix to investigate the role of Si addition in the matrix on the radiation stability of the phase(s) in the U–7Mo fuel/matrix interaction layer. A similar interaction layer that forms in irradiated U–7Mo dispersion fuels with pure Al matrix has been found to exhibit poor irradiation stability, likely as a result of poor fission gas retention. The interaction layer for both U–7Mo/Al–2Si and U–7Mo/Al fuels is observed to be amorphous. However, unlike the latter, the amorphous layer for the former was found to effectively retain fission gases in areas with high Si concentration. When the Si concentration becomes relatively low, the fission gas bubbles agglomerate into fewer large pores. Within the U–7Mo fuel particles, a bubble superlattice ordered as fcc structure and oriented parallel to the bcc metal lattice was observed where the average bubble size and the superlattice constant are 3.5 nm and 11.5 nm, respectively. The estimated fission gas inventory in the bubble superlattice correlates well with the fission density in the fuel.     

Phase relations in the U-Mo-Al ternary system

The phase relations in the U-Mo-Al system of quenched samples annealed at 800 °C for 2 weeks and at 400 °C for 2 months have been established using X-ray powder diffraction, scanning electron microscopy and energy dispersive spectroscopic analysis performed at room temperature. Two ternary Al-rich phases, UMo_2−xAl_20+x and U₆Mo_4+xAl_43−x are found stable at 800 °C and 400 °C. They show significant homogeneity ranges resulting from Mo/Al substitution mechanism on various mixed crystallographic sites, as evidenced by single-crystal structure refinements. Substitution of up to 25 at.% of Al by Mo atoms is also observed for UAl₂ (cubic MgCu₂-type) giving a quite large extension (UAl_2−xMo_x, 0 < x < 0.5) into the ternary system. Larger substitution (0.6 < x < 0.7 at T = 800 °C) stabilizes another ternary Laves phase, UAl_2−xMo_x with the hexagonal MgZn₂-type. There is no detectable solubility of Mo in UAl₄, and it is of the order of 1 at.% in UAl₃. The interaction layers between the γU-Mo alloys and the Al matrix in nuclear fuel plates can be successively estimated as composed of the two- and three-phase fields equilibrium indicated on the assessment of the phase relations drawn for samples heat-treated at 400 °C.     

Ionization in symmetric and nearly symmetric low energy ion-surface collisions

Multiply charged ions are emitted following bombardment of Al(1 0 0) and Si(1 1 1) by low energy Si⁺ and P⁺ ions. The ion formation is attributed to inner-shell electron promotion during a hard collision between symmetric or nearly symmetric atomic species, followed by Auger decay outside the surface. The relative yield of triply charged Si ions for Si⁺ → Si(1 1 1) is much smaller than that of triply charged Al ions in direct recoil Si⁺ → Al(1 0 0) experiments. This difference can be explained by assuming that only one 2p hole is produced in a Si atom during the symmetric collision, whereas a double 2p hole is also produced in the Al atom following the nearly symmetric Si-Al collision. Further evidence is provided by the complimentary experiment P⁺ → Si(1 1 1), where Si³⁺ regains its intensity and Si⁴⁺ emerges as a result of a double 2p hole decay with shake-off.     

Physical properties of monolithic U8 wt.%-Mo

As a possible high density fuel for research reactors, monolithic U8 wt.%-Mo (“U8Mo”) was examined with regard to its structural, thermal and electric properties. X-ray diffraction by the Bragg-Brentano method was used to reveal the tetragonal lattice structure of rolled U8Mo. The specific heat capacity of cast U8Mo was determined by differential scanning calorimetry, its thermal diffusivity was measured by the laser flash method and its mass density by Archimedes’ principle. From these results, the thermal conductivity of U8Mo in the temperature range from 40 °C to 250 °C was calculated; in the measured temperature range, it is in good accordance with literature data for UMo with 8 and 9 wt.% Mo, is higher than for 10 wt.% Mo and lower than for 5 wt.% Mo. The electric conductivity of rolled and cast U8Mo was measured by a four-wire method and the electron based part of the thermal conductivity calculated by the Wiedemann-Frantz law. Rolled and cast U8Mo was irradiated at about 150 °C with 80 MeV ¹²⁷I ions to receive the same iodine ion density in the damage peak region as the fission product density in the fuel of a typical high flux reactor after the targeted nuclear burn-up. XRD analysis of irradiated U8Mo showed a change of the lattice parameters as well as the creation of UO₂ in the superficial sample regions; however, a phase change by irradiation was not observed. The determination of the electron based part of the thermal conductivity of the irradiated samples failed due to high measurement errors which are caused by the low thickness of the damage region in the ion irradiated samples.     

The interdiffusion and solid-state reaction of low-energy copper ions implanted in silicon

At room temperature, single-crystal silicon was implanted with Cu⁺ ions at an energy of 80 keV using two doses of 5 × 10¹⁵ and 1 × 10¹⁷ Cu⁺ cm⁻². The samples were heat treated by conventional thermal annealing at different temperatures: 200 °C, 230 °C, 350 °C, 450 °C and 500 °C. The interdiffusion and solid-state reactions between the as-implanted samples and the as-annealed samples were investigated by means of Rutherford backscattering spectrometry (RBS) and X-ray diffraction (XRD). After annealing at 230 °C, the XRD results of the samples (subject to two different doses) showed formation of Cu₃Si. According to RBS, the interdiffusion between Cu and Si atoms after annealing was very insignificant. The reason may be that the formation of Cu₃Si after annealing at 230 °C suppressed further interdiffusion between Si and Cu atoms.     

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.     

Corrosion experiments and materials developed for the Japanese HLM systems

The static corrosion tests in lead-bismuth eutectic (LBE) were conducted from 450 °C to 600 °C to understand corrosion behavior and develop corrosion resistant materials for heavy liquid metal systems. While increase of Cr content in steels enhances corrosion resistance in LBE, the effect approaches a constant value above 12 wt% of Cr. Corrosion depth in LBE increases with increasing temperature and corrosion attack becomes severe above 550 °C even under the condition of high oxygen concentration. Nickel dissolution and Pb-Bi penetration occur in 316SS and JPCA above 550 °C under the condition of high oxygen concentration. When oxygen concentration decreases below the level of Fe oxide formation, corrosion attack on these steels also becomes violent due to dissolution of various elements and grain boundary corrosion. Whereas additions of 1.5 wt% Si to T91 and 2.5 wt% Si to 316SS improve corrosion resistance, the effect is insufficient taking fluctuation of oxygen concentration in LBE into consideration. Furthermore, addition of 1.5 wt% Si to T91 causes rise in DBTT. A new coating method using Al, Ti and Fe powders produces corrosion resistant coating layers on 316SS. The coating layers containing 6-8 wt% Al exhibit good corrosion resistance at 550 °C for 3000 h in LBE containing 10⁻⁶-10⁻⁴ wt% of oxygen.