Light emission from carbon-based materials under ITER relevant thermal shock loads

Light emission from carbon-based materials (fine grain graphite, CFC and silicon doped CFC) was observed during ITER relevant thermal shock loads by means of in situ optical diagnostics. The light emission which corresponds to particle release clearly indicated different particle release processes in the three materials. The differences were also found in the initiation temperatures of particle release and the surface morphology of the loaded areas. These results are related to the thermal stress in bulk materials. In addition to particle release, vapor cloud formation caused by thermal shock loads were observed as CII lines and lines from the C₂ Swan system. No Si lines but lines from SiC₂ molecules (Merrill-Sanford bands) were observed in Si doped CFC. This indicates that atomic silicon is not released under ITER relevant thermal shock loads.     

Luminescence properties of zirconium oxide films

Various luminescence techniques have been applied in the characterisation of the zirconium oxide film formed on the metal matrix. Our investigation shows that some alloying elements in the zirconium alloys show characteristic emissions which could be used to study their distribution in the zirconium oxide. While the origin and detailed mechanism of luminescence in the zirconium oxide system is not well established, cathodoluminescence (CL) and thermoluminescence results from this study support the theory that oxygen vacancies in complexes with impurities are responsible for the intrinsic luminescence in the oxide system. The specific oxygen vacancy and impurity complex is reported in literature as the T-defect, having an energy of 2.2 eV. To aid in interpretation of the CL data, optical transmission properties of the oxides were also measured. The latter investigation showed the thickness of the oxide illuminated by the CL technique is greatly dependent on the optical transmission properties of the oxide, and it ranges from 20 μm for pure zirconium or low-alloyed zirconium oxides to less than 3 μm for oxides of alloys with more absorbing, or higher alloying element concentrations. The 20 μm depth of illumination is significantly deeper than the electron penetration depth, which suggests a secondary source of excitation, possibly characteristic X-rays emitted by the specimen. These combined properties of the oxide and technique can result in CL images showing structural contrast not easily seen during SEM or optical microscopy of oxidised surfaces.     

On the solubility of chromium sesquioxide in uranium dioxide fuel

Chromium sesquioxide doped uranium dioxide systems with a variety of initially added Cr₂O₃ amounts were prepared under different sintering conditions. The solubility limit of Cr for each sintering condition is derived from the measured content of the dopant dissolved in the UO₂ matrix using electron probe microanalysis (EPMA). It is found that the solubility of chromium in a uranium dioxide system for these series sintered at 1600, 1660 and 1760 °C is limited to respectively 0.065±0.002, 0.086±0.003 and 0.102±0.004 wt% Cr. The lattice parameters of the different Cr₂O₃ doped fuels have been examined with X-ray diffraction (XRD). The XRD peaks of the samples sintered at 1760 °C as well as a UO₂ reference without Cr, prepared under the same conditions, were measured and a value for the lattice parameter a for each sample was obtained using the unit cell refinement method. A slight contraction of the lattice parameter is observed with increasing dopant content.     

Water formation in graphite and boron-doped graphite under simultaneous O and H irradiation

When graphite is exposed to simultaneous irradiation by H⁺ and O⁺, in addition to the H⁺-induced hydrocarbon and O⁺-induced CO and CO₂ formation, water is also formed. The present investigation explores the effect of the presence of boron in graphite on water formation. The results show that B-doped graphite specimens (∼15 at.% B) exposed to simultaneous O⁺ and H⁺ irradiation produce less water when compared to pure graphite under similar irradiation. The concept of the formation of ‘water-precursors’ at the end of the O⁺ range is proposed to explain the observed effect.     

Phase relations in the quaternary Fe-Pu-U-Zr system

The phase diagram in the quaternary Fe-Pu-U-Zr system was established at 923 K in the uranium-rich region to understand better the compatibility between the metal fuel and stainless steel cladding in a fast reactor. The experimental phase relation data obtained in this study was applied to the thermodynamic methodology for construction of the phase diagram. The calculated phase diagram was consistent and well within the experimental data. The applicability of the thermodynamic model to other temperatures was confirmed by comparing the present results of differential thermal analysis with the calculated phase diagrams. These consistencies mean that both the thermodynamic model and the assessed parameters in the binary and ternary subsystems developed so far are reasonable. The calculated phase diagram established in this study was also in good agreement with the analysis of the diffusion zone in tests on Pu-U-Zr/Fe couples. This suggests that the diffusion zone formed at the fuel-cladding interface in the reactor system can be assessed using the phase diagram in the quaternary Fe-Pu-U-Zr system.     

Elastic and electronic properties and stability of SrThO3, SrZrO3 and ThO2 from first principles

First-principle calculations in the framework of the full-potential linearized-augmented-plane-wave method (FLAPW, as implemented into the WIEN-2k code) have been performed to understand the structural, elastic, cohesive and electronic properties of the meta-stable cubic strontium thorate SrThO₃. The optimized lattice parameters, elastic parameters, formation energies, densities of states, band structures and charge density distributions are obtained and discussed in comparison with those of cubic SrZrO₃ and ThO₂.     

Experimental studies of mechanical properties of solid zirconium hydrides

Zirconium alloy Zr-2.5Nb has been hydrided to ZrH_x (x = 1.15-2.0), and studied using microhardness and unconfined and confined compression techniques. At room temperature, results on Young’s modulus and yield strength of solid hydrides show that these mechanical properties remain about the same as the original zirconium alloy for hydrogen compositions up to about ZrH_1.5. The levels of these properties start to drop when δ hydride becomes the major phase and reaches a minimum for the ε hydride phase. Between room temperature and 300 °C, Young’s modulus of solid hydrides decreases with temperature at about the same rate as it does for the original zirconium alloy.     

Influence of annealing in air on transport properties of long-time oxidized layers of zirconium alloys

The change of resistivity and permittivity of oxide films grown 1554 days in water at 360 °C on tubes of Zr1Nb, ZrSnNb(Fe) and IMP Zry-4 was investigated by I-V measurements after long-time annealing in air at 123 °C. Thereby the room temperature resistivity, highest in Zry-4W and lowest in Zr1Nb, increased by more than two orders of magnitude due to a strong decrease of mobility, with nearly constant carrier concentration. The polarity of the zero current changed from negative to positive values in the pA range. The changes are thought to be due to continued oxidation.     

Disproportionation of Pu(IV): A reassessment of kinetic and equilibrium properties

Kinetic results for disproportionation of Pu(VI) and reaction of Pu(III) with Pu(VI) show that rates of in HNO₃ and HClO₄ solutions are described by trimolecular rate laws consistent with involvement of trimeric hydroxo complexes as reactive intermediates in the slow mechanistic steps. Product ratios and modeling of concentration-time curves reveal that Pu(V) is formed by reduction of Pu(VI) product in a secondary reaction. Results do not support the accepted interpretation that attributes reversible reaction and Pu(V) formation to a two-step bimolecular process. Secondary redox reactions driven by disproportionation of Pu(VI) prevent attainment of equilibrium in 1 M H⁺ and determine long-term redox chemistry. The equilibrium constant (0.00051) defined by forward and reverse rate constants for 1 M HClO₄ agrees with that (0.00049) derived from concentration data for 1 M HNO₃, but not with prior results. Disagreement of these values with that calculated from thermodynamic data suggests that steady-state Pu concentrations are controlled by kinetics. Possible pathways of secondary reactions are identified and a mechanism for reversible oxygenation of plutonium ions is described.     

Photoluminescence study of swift heavy ion (SHI) induced defect centers in sapphire

Single crystals of sapphire (Al₂O₃: Fe, Ti, Cr) were irradiated at room temperature with different fluence of 100 MeV Ni ions. Photoluminescence (PL) spectra of pristine and irradiated sapphires were recorded at room temperature under 2.8 eV blue excitation. A broad emission band consists of two bands centered at 516 nm corresponding to F₂ defect center and 546 nm corresponding to defect center was observed. The intensity of these defect centers was found to vary with the fluence. defect center develops at low fluence reaching maximum at 5 × 10¹⁶ ions/m² and finally decreasing at higher fluence. The behavior is interpreted in terms of creation of defect centers, their clustering and annihilation.     

Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.     

Analysis for hydrogen particle balance of plasma-wall system in the large helical device

The hydrogen particle balance of the plasma-wall system in the large helical device (LHD) is analyzed, using a zero dimensional model for plasma particles, neutrals in vessel and hydrogen inventory in wall. Based on the measurement of neutral gas pressure, plasma density and the pumping speed of the cryo-pumping system, it is found that the hydrogen retained in the wall desorbes with short and long time constant. The short term desorption is of order of 10²¹ atoms with a time constant of a few minutes, which is much smaller than the wall pumping for one shot, 10²² atoms. In a long time scale of about one experimental day, the wall absorbs significantly large amounts of hydrogen, up to 10²⁴ atoms. One of the possible reasons for the large wall pumping is a carbon deposition layer on the first wall surface. The effect of hydrogen retention on density control is also discussed.     

Chamber wall materials response to pulsed ions at power-plant level fluences

Candidate dry-wall materials for the reactor chambers of future laser-driven Inertial Fusion Energy (IFE) power plants have been exposed to ion pulses from RHEPP-1, located at Sandia National Laboratories. These pulses simulate the MeV-level ion pulses with fluences of up to 20 J/cm² that can be expected to impinge on the first wall of such future plants. Various forms of tungsten and tungsten alloy were subjected to up to 1600 pulses, usually while being heated to 600 °C. Other metals were exposed as well. Thresholds for roughening and material removal, and evolution of surface morphology were measured and compared with code predictions for materials response. Powder-metallurgy (PM) tungsten is observed to undergo surface roughening and subsurface crack formation that evolves over hundreds of pulses, and which can occur both below and above the melt threshold. This roughening is worse than for other metals, and worse than for either tungsten alloyed with rhenium (W25Re), or for CVD and single-crystal forms of tungsten. Carbon, particularly the form used in composite material, appears to suffer material loss well below its sublimation point. Some engineered materials were also investigated. It appears that some modification to PM tungsten is required for its successful use in a reactor environment.     

Improvements in performance and microstructure of doped graphites as plasma-facing materials by a new process

Multi-elemental (B, Si and Ti) doped graphite was prepared from fillers of petroleum coke, dopants and binder of coal tar pitch by a new liquid mixing method. Such composite has not only high thermal conductivity (233 W/m K) but also excellent mechanical strength compared with the material prepared by the conventional solid mixing method. Scanning electron microscopy (SEM) results show that such material has a fine-grained structure as well as little pore diameter. The effects of the manufacturing procedure of doped graphites on the performance and microstructure are investigated. In addition, correlations between properties and microstructure are also discussed in detail.     

Tritium management in the first-wall materials of A-DC and TAURO blankets

The tritium management in the first wall of two European breeding blanket options, A-DC and TAURO, has been simulated numerically to analyse the influence of the material selected: ODS-RAFM steel for the Advanced Dual-Coolant (A-DC) and SiC_f/SiC composite for TAURO options. The SRIM code has been used to simulate triton implantation and define the tritium source in each kind of material as a function of the depth. The TMAP4 code was used to analyse the posterior transitory gas transport process within the material, while taking into account the tritium transport properties of each material and the temperature variation through material thickness and operating time. Both the transient evolution and the final steady-state tritium transport behaviour have been characterised. The tritium transient flux to the coolant, the recycling flux and the absorbed tritium transient inventories have been simulated. Main conclusions have been drawn about the tritium performance of each first wall.     

Effects of natural aging on the tensile properties of water-quenched U-6%Nb alloy

Uranium-6 wt% niobium (U-6%Nb) alloy has been in use for many years in the water-quenched (WQ) condition. The purpose of this work was to determine the effect of natural aging on tensile properties of the WQ U-6%Nb alloy. The materials studied were hemispherical shells after 15 and 20 years in storage. The alloy was successfully tested in the original curved configuration, using the specially designed tensile test apparatus. Finite element analysis confirmed the validity of the test method. The results of the tensile tests clearly indicated that in the WQ condition, the material is changing and after 15 and 20 years, the yield strength exceeds the original maximum allowable specification. The fracture mode transitions from highly ductile, microvoid coalescence in new material to a mixed mode of shallow dimples and inclusion-induced voids in the naturally aged material.     

Ab initio formation energies of Fe-Cr alloys

We have calculated ab initio lattice parameters, formation energies, bulk moduli and magnetic moments of Fe-Cr alloys. The results agree well with available experimental data. In addition to body centered cubic (bcc) alloys, which are representative of ferritic steels used in fast neutron reactors, face centered cubic (fcc) and hexagonal close packed (hcp) phases were considered in order to complete a theoretical database of thermodynamic properties. Calculations were done for the ferromagnetic phase, as well as for a phase with local moment disorder, simulating the magnetic structure at high temperatures. For the latter case, the formation energy of the alloy is strictly positive smooth function of chromium concentration, in agreement with experiments performed at high temperature. In the ferromagnetic case, a negative mixing enthalpy is found for chromium concentrations below 6%. Our observation is consistent with the experimentally observed inversion of the ordering trend, as well as with formation of the chromium rich α^′ phase at Cr-concentrations above 9%, occurring at T<900 K.     

Fluence rate effect semi-mechanistic modelling on WWER-type RPV welds

Effort at JRC-IE is ongoing in order to develop a semi-mechanistic model to forecast radiation embrittlement. The understanding and the quantification of the influence of the fluence rate is of particular importance for the correct interpretation of data obtained in material testing reactors or in surveillance capsules, which are accelerated with respect to embrittlement of the reactor pressure vessel wall itself. To verify the applicability of the fluence rate as included in the semi-mechanistic model and tuning the model parameters various WWER-type vessel weld material have been studied. For the selected welds, copper ranges from 0.08 to 0.18 mass%, while phosphorus variation is from 0.013 to 0.036 mass%. The fluence range is up to 2 × 10²⁰ n cm⁻² obtained at two fluence rates of 4 × 10¹¹ and 3.5 × 10¹² n cm⁻² s⁻¹, typical for WWER-440 surveillance positions. Significant fluence rate effect has been observed for the welds containing low copper and moderate phosphorus, and adaptation of the semi-mechanistic model’s parameters for the high flux data is required. To verify the consistency and the limits of the findings other similar data coming from RPV surveillance programmes are also included in this analysis.     

Simulation study of carbon impurity dynamics on tungsten surfaces exposed to hydrogen ions

The depth profile of C impurity deposited on a W target exposed to H⁺ and C⁺ impurities at a concentration of C: 0.8% has been calculated in terms of segregation, diffusion and chemical erosion. For the segregation, the Gibbsian model has been used. For the diffusion, a concentration dependent diffusion model (C in WC and/or C) has been utilized. For the chemical erosion, the chemical erosion yield much lower than that for the H-C system has been applied. The calculated depth profiles at 653 K and 913 K are in good agreement with the XPS data. The agreement indicates that there is a significant contribution of segregation, which shifts the maximum C concentration to the top surface in the depth profiles. On the other hand, there are little contributions from diffusion and chemical erosion, which are related closely to formation of WC in the target.     

Reversible expansion of gallium-stabilized δ-plutonium

It is shown that the transient expansion of plutonium-gallium alloys observed both in the lattice parameter as well as in the dimension of a sample held at ambient temperature can be explained by assuming incipient precipitation of Pu₃Ga. However, this ordered ζ′-phase is also subject to radiation-induced disordering. As a result, the gallium-stabilized δ-phase, being metastable at ambient temperature, is driven towards thermodynamic equilibrium by radiation-enhanced diffusion of gallium and at the same time reverted back to its metastable state by radiation-induced disordering. A steady state is reached in which only a modest fraction of the gallium present is arranged in ordered ζ′-phase regions.