Journal of Nuclear Materials最新文献

Hydrogen embrittlement is a crucial factor for the performance and life of zirconium alloys in nuclear industry. During service, the residual stress in the materials induces re-distribution of hydrogen, which further affects the fracture behavior. This study is dedicated to investigating the hydrogen diffusion and precipitation under the meso‑scale non-uniform deformation field. The texture effect of hydrogen behavior was examined by considering three different texture scenarios: grains with c-axis deviating from the tensile direction of random angle, 0°and 90° The cases were termed Random, T000, and T0900 respectively. The crystal plasticity finite element method (CPFEM) was employed to simulate the stress-assisted diffusion of hydrogen atoms in the polycrystalline zirconium alloy. It is determined that the T0900 case gets fewer regions with higher hydrostatic stress gradient, which helps relieve the hydrogen concentration. Compared to the T0900 texture, the interaction between soft and hard grains in the Random and T000 texture conditions results in higher stress gradient and severe hydrogen concentration. Statistical analysis was conducted and the hydrogen concentration in the three textures presents a normal distribution. T0900 gets a relatively lower overall hydrogen concentration while the T000 texture gets severe hydrogen concentration larger than 120 wt.ppm. In Random and T000 textures, hydrogen tends to accumulate at grain interior and boundaries, and continuous hydrogen concentration band forms along adjacent GBs. The T0900 scenario is free of large continuous hydrogen concentration zones, which helps reduces the adverse effects of hydrogen diffusion and precipitation. The findings of the work advance the understanding of the hydrogen behavior affected by stress heterogeneity and texture, which is the basis for the exploration of hydrogen embrittlement.

Oxide dispersion strengthened Cu (ODS Cu) alloys possess a high potential to be used as promising heat sink materials for the divertor system in DEMO fusion reactors. Considering their future application in fusion reactor divertor systems, mass-production of ODS Cu alloys is a must. The purpose of this study is to investigate the feasibility of large-scale production of ODS Cu alloys by improving the production process based on the already established mass-production process of ODS Fe alloys in cooperation with the steel manufacturer Kobelco Research Institute. After optimization of the mechanical alloying (MA) parameters, a large-scale production of ODS Cu powders was successfully achieved with a recovery rate of over 90% and around 1Kg powder in one batch production using an industry-level attritor ball mill cooled by circulating water. The mass-produced ODS Cu alloys were comprehensively characterized, including mechanical properties, microstructure and thermal diffusivity, and a new concept of low-energy ball milling was proposed based on the results to realize the mass-production of ODS Cu alloys.

This study investigated the solubility of UO₂ in LiCl-2 wt% Li₂O at 650 °C with O₂ partial pressures up to 114 torr. Experiments were run in which UO₂ powder was contacted with the molten salt for up to 120 h. Salt samples were taken and analyzed for U and corrosion product concentration using inductively coupled plasma mass spectrometry and for Li₂O concentration using titration. The highest measured U concentration in the salt was 0.06 wt%, and O₂ partial pressure was observed to have little to no effect on the solubility. Increasing O₂ partial pressure did increase the concentration of corrosion products in the salt. The concentration of Li₂O in the molten salt progressively decreased with time in contact with UO₂. This can be explained by the formation of Li₂UO₄ via reaction of UO₂ with Li₂O.

The pseudo-binary metallic fuel alloy, U-10M (wt%, M is the optimal combination of Mo, Ti, and Zr), has the potential to increase fuel solidus temperature, reduce the onset temperature of body-centered cubic phase, and increase the fuel's chemical stability compared with the conventional U-10Zr (wt%) metallic fuel. Post Irradiation Examination (PIE) confirmed excellent fuel performance for the U-10M (10M=5Mo-4.3Ti-0.7Zr wt%) fuel irradiated in the Advanced Test Reactor (ATR) to 2.2 at% burnup at Peak Inner Cladding Temperature (PICT) of 650 °C, an upper bound temperature for metallic fuel. But previous PIE study also observed Fuel Cladding Chemical Interaction (FCCI) on the cladding side, which is known as a fuel performance limiting issue but has not been fully understood yet. As an effort to improve the understanding of FCCI phenomenon, this study performed Scanning Electron Microscopy (SEM) and in-depth Transmission Electron Microscopy (TEM) characterization on a FCCI region. The examined FCCI region is dominated by (U, Zr)(Fe, Cr)_2, suggesting that U-Fe interdiffusion reaction played a key role in inducing FCCI. Additionally, the FCCI boundary into the cladding consists of four distinctive phases, (U, Zr)(Fe, Cr)₂, fcc-Cr, tetragonal UCr_0.1Fe_9.9Si₂, intermetallic σ-FeCr, and lanthanide fission products at concentration up to ∼5.5 at%. Another goal of this study is to verify the involvement of Ti and Mo in FCCI formation on the cladding side. Observable Ti is found halfway of the thickness in the examined FCCI region, while 0.3–5 at% Mo is detected across the entire thickness of the examined FCCI region. Neither Ti nor Mo reacted with HT9 cladding constituents despite their diffusion footmark. Therefore, alloying Ti and Mo into the U-Zr fuel should not complicate the interdiffusion reaction on the HT9 cladding side.

IFMIF-DONES is a powerful neutron source which is being designed with the main purpose of qualifying structural materials (EUROFER being the first candidate) to be used in DEMO and fusion power plants envisaged after it. This source relies on one high current (125 mA) deuterons beam accelerated to 40 MeV which impacts on a liquid lithium target to produce an intense neutron flux through Li(d,n) stripping reactions able to simulate the nuclear responses expected on the first wall of the reactor. The engineering design of IFMIF-DONES is presently being carried out mainly in the framework of the EUROfusion Work Package Early Neutron Source (WPENS). Since 2021, a new phase has started with the launch of the FP9 WPENS workplan whose objective is to continue advancing the engineering design of the facility, putting special effort on the experimental validation of those aspects which still need to be qualified to demonstrate the fulfillment of functional and safety requirements. The ENEA Brasimone Research Centre has been and still is strongly committed in several validation tasks concerned in particular with the lithium systems design and the Remote Handling (RH) maintenance.

In this paper, an overview of the most relevant validation activities carried out in recent years or still in progress or planned at the ENEA Brasimone Research Centre in both of the aforementioned areas is presented, including the erosion/corrosion tests in the Lifus 6 loop; the nitrogen-gettering materials qualification in the ANGEL facility; and the RH and prototypes testing in the DRP laboratory for the validation of the High Flux Test Module electric connectors and the pre-heating of the Target Assembly.

Two tests, AN3 and AN4 from the RISOE-III Fission Gas Release (FGR) project are analysed using the updated version of the FALCON MOD01 code coupled with the GRSW-A model. Based on comparison of the stand-alone code calculation with the measurement for centerline fuel temperature, the previous calibration and verification of the thermal analysis of the code is confirmed. Additional analysis with the FALCON-to-FRELAX (F2F) coupled code system is carried out and presented for simulation of the effects of delayed axial redistribution of the released fission gases in the rodlets, as an attempt to interpret the data on significant jumps in measured upper-plenum pressure at a power dip. The other feature of the data, addressed by the additional sensitivity study with F2F, is peaking that can be seen in the measured fuel temperature after each power rise. A conclusion can be proposed that the peculiarities in the online-measurement could, probably, result from the features of fabrication of the fuel segments for pre-irradiation in the NPP, and instrumentation of the rodlets to be tested in the research reactor. Therefore, these features can have only limited effect on evaluation of reliability and safety of the standard fuel during normal operation in power reactors.

It is significantly important to improve the corrosion resistance of Zr-Sn-Nb alloys in dissolved oxygen (DO) high temperature water by compositional modification. Here, the effect of Cu addition on the corrosion behaviors of Zr-0.5Sn-0.2Nb-0.4Fe-0.2Cr alloy was studied by using an autoclave circulating water loop at 360 °C/20 MPa up to 250 days. The oxide microstructures were characterized by a combination of SEM, TEM, STEM-EELS and PED. The results showed that compared with 0Cu alloy and 0.07Cu alloy, 0.13Cu alloy shows better corrosion resistance and its final weight gain after 250 days exposure is even lower than that of commercial Zr-4 alloy. The oxides for 0.07Cu and 0.13Cu alloys show more ordered crystals, thinner oxide and less lateral cracks, and the columnar grains show a stronger {10–3} texture, compared to Cu-free counterpart. The t-Zr₂Cu SPPs have oxidized in the oxygen-rich layer below the O/M interface. The Cu ions tend to diffuse towards the grain boundary of dense columnar grains and maybe strengthen the grain boundaries, avoiding the defect generation under intrinsic stress in the oxide.

The corrosion behaviour of Al-containing alloys in KCl-MgCl₂ at 700 °C was investigated and compared with that of commercial alloys. The developed Al-containing alloys showed less or similar weight loss compared to Hastelloy N owing to Al-rich oxide formation. While 316 SS showed both uniform and localised corrosion, a high Al-containing alloy was almost unattacked by any form of corrosion. The SEM map in this study innovatively showed and confirmed the high Cr depletion at the top matrix layer of the exposed 316 SS alloy. It was found that the primary controlling mechanism of corrosion is the outward diffusion of metal ions promoted by the volatile metal chlorides, while Cr diffusion was detrimental, Al diffusion was beneficial in mitigating corrosion via the formation of Al oxides. Noteworthily, this study revealed that the KCl-MgCl₂ salt was very corrosive to the 316 SS, and its interaction led to an obvious intergranular corrosion arising from Cr-depletion of the grain boundaries due to the high Cr content of the 316 SS. While for the high Al-containing alloys, the ACES with lower Al content and higher Ni content exhibited a higher corrosion rate when in contact with KCl-MgCl₂salt than the ADSS with higher Al content and lower Ni content. The mechanism of electrochemical interaction between the KCl-MgCl₂ salt and the high Al-containing alloys was that, the increased Al content and decreased Ni content enabled the formation of α – Al₂O₃, which provided excellent protective barrier against corrosion attack, while, the lower Al and higher Ni contents lead to the formation of Mg and Ni aluminate oxides. These oxides were less protective against corrosion attack in the molten salt environment.

To investigate the coupling between nuclear and electronic energy losses in UO₂, we irradiated thin foils with 0.39 MeV Xe and/or 6 MeV Si ions at 93 K using single or simultaneous dual beam ion irradiations. The evolution of perfect dislocation loops was characterized by in situ transmission electron microscopy (TEM). Additional ex situ TEM characterizations at room temperature revealed for the first time in UO₂ the presence of faulted Frank loops too small to be measured during in situ experiments and conventional bright field kinematical imaging conditions.

For the single Xe irradiation, which favor dominant ballistic energy losses, we observed a continuous nucleation of small perfect dislocation loops, which increase in size for our last fluences by growing through mainly coalescence effect. Both the single Si and dual Xe & Si irradiations showed a coupling between nuclear and electronic energy losses, resulting in a significant loop density increase and a tangled line network formation, respectively. These phenomena occur at lower dpa levels, compared to the single Xe irradiation, likely resulting from the thermal spike effect of Si ions. The present results were compared to our previous work at 293 K to investigate the role of irradiation temperature on the energy losses coupling. For the Xe irradiation, the density increases and the loops are smaller at 93 K compared to 293 K, resulting from the uranium interstitials mobility being prevented or allowed. For the Si irradiation, the dislocation evolution kinetics are similar at both temperatures. The electronic excitations effect seems greater than the irradiation temperature effect in this temperature range. For the Xe & Si irradiation, the loop kinetics change resulting in a tangled line network formation is faster and thus the loop transformation into lines occurs at lower dpa levels at 93 K compared to 293 K. It appears that the irradiation temperature affecting the mobility of some small point defects reduces the electronic excitation effect in this case.

A comprehensive study by integrating first principles molecular dynamic (FPMD) simulations and differential scanning calorimetry (DSC) experiments on the physicochemical properties of multiple halophilic fission products (MF_n = RbF, SrF₂, YF₃, and ZrF₄) and molten NaF-BeF₂ (FNaBe) mixed salts is reported. The effect mechanism of product type on density and specific heat capacity has been discussed from the interionic distance and its stability, coordination number and its distribution, as well as the neighbor cluster structure and its orientation. It is concluded that volume increments of molten FNaBe+MF_n are primarily caused by Na-Na and M-F interactions, while the highest specific heat capacities (c_p) of molten FNaBe+ZrF₄ is closely related to the stability of Zr-F bonds and its coordination structures. Furthermore, the DSC results assisted by quantitative analysis of elements indicate that the additive concentration of YF₃ is positively correlated with c_p of molten FNaBe+YF₃ in a certain range.