ALBERT

Description: 〈p〉Publication date: October 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 524〈/p〉 〈p〉Author(s): William E. Frazier, Shenyang Hu, Douglas E. Burkes, Benjamin W. Beeler〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Experiments show that irradiation-induced recrystallization speeds up the swelling kinetics in U-10 wt% Mo fuels. However, recrystallization mechanisms and the effect of initial grain microstructures on recrystallization kinetics are still unclear. In this work a Monte Carlo model coupling the rate theory of defect evolution has been developed to study the irradiation-induced recrystallization. The rate theory is used to describe the spatial evolution of gas bubbles, interstitials and interstitial loops; First-Passage Kinetic Monte Carlo (FPKMC) approach is used to describe the fast and strongly anisotropic migration of interstitials, and a Cellular Automata method is used to model the formation of recrystallized grains. With the assumption that recrystallization may occur when the local interstitial loop density is larger than a given critical value, simulation results reveal that 1) recrystallized grains first nucleate on grain boundaries and the recrystallization zone front moves to the center of original coarse grains in the UMo matrix, and 2) recrystallization starts earlier in coarse polycrystalline structures, while the overall recrystallization kinetics decreases with increasing grain size. These results agree with experimental observations. The comparison of recrystallization kinetics obtained from experiments and modeling suggests that the interstitial loop accumulation leads to the recrystallization and the interstitial loop growth is suppressed inside coarse grains due to the over-pressured intra-granular gas bubbles.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: October 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 524〈/p〉 〈p〉Author(s): Hao Shi, Adrian Jianu, Alfons Weisenburger, Chongchong Tang, Annette Heinzel, Renate Fetzer, Fabian Lang, Robert Stieglitz, Georg Müller〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉 〈p〉The goal of the study was to determine the critical concentrations of Al, Cr and Ni, at which the quaternary Fe–Cr–Al–Ni model alloys, exposed to oxygen-containing molten Pb up to 600 °C, are corrosion resistant, while preserving the austenite structure of the alloy matrix.〈/p〉 〈p〉Twelve alloys were designed to meet the above mentioned requirements; six of them showed corrosion resistance and preserved the austenite phase in the alloy bulk, during the exposure at 550 °C and 600 °C for 1000 h to molten Pb containing 10〈sup〉−6〈/sup〉 wt% oxygen. Based on experimental results a general formula was substantiated as follows: Fe-(20–29)Ni-(15.2–16.5)Cr-(2.3–4.3)Al (wt.%). In case of temperatures below 550 °C, the critical Cr content was 14.4 wt%.〈/p〉 〈p〉Two corundum-type crystalline structures were identified as the constituent phases of the passivating scales, one being Cr〈sub〉2〈/sub〉O〈sub〉3〈/sub〉 and the other Al〈sub〉2〈/sub〉O〈sub〉3〈/sub〉–Cr〈sub〉2〈/sub〉O〈sub〉3〈/sub〉 solid solution. The average amount of Cr〈sub〉2〈/sub〉O〈sub〉3〈/sub〉 in the Al〈sub〉2〈/sub〉O〈sub〉3〈/sub〉–Cr〈sub〉2〈/sub〉O〈sub〉3〈/sub〉 solid solution, found in the passivating scales of the Fe–Cr–Al–Ni model alloys, was estimated at ≈ 40 wt% at 550 °C and ≈35 wt% at 600 °C.〈/p〉 〈p〉A transitional layer, consisting of Fe- and Ni-enriched austenitic matrix and exhibiting randomly distributed intermetallic B2-(Ni,Fe)Al, was formed below the oxide scale up to a depth of two microns.〈/p〉 〈p〉The austenite, as matrix, and Ni〈sub〉3〈/sub〉(Al,Fe) as precipitates are the microstructural phases of the bulk alloys after exposure for 1000 h at 600 °C to oxygen-containing molten Pb.〈/p〉 〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311519301394-fx1.jpg" width="485" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: October 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 524〈/p〉 〈p〉Author(s): Jo Jo Lee, Stephen S. Raiman, Yutai Katoh, Takaaki Koyanagi, Cristian I. Contescu, Xunxiang Hu, Ying Yang〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Silicon carbide is widely appreciated for its high temperature strength, radiation tolerance and neutronic transparency in applications for fuel particles and core internals of nuclear reactors. In the Fluoride Salt-Cooled High Temperature Reactor, silicon carbide ceramic matrix composites are candidate construction material for regions of higher neutron fluxes. Silicon carbide is wettable and reacts electrochemically with dissolved metals. Metallic impurities, tritium, moisture-based impurities and fission products, as well as thermal gradients can accelerate hot corrosion of silicon carbide in molten fluoride salt. Tritium can become trapped in radiation defects of silicon carbide. Thus, an understanding of the potential for tritium absorption, impurities reactions and thermal gradient-assisted corrosion mechanisms along with tritium recovery and redox control systems are necessary to mitigate silicon carbide corrosion in molten fluoride salt systems. Here, we survey current research on silicon carbide corrosion in molten fluoride salts and critically evaluate the research and development gaps.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 21 August 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Li-Fang Wang, Bo Sun, Hai-Feng Liu, De-Ye Lin, Hai-Feng Song〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Thermodynamic and kinetic properties of intrinsic point defects in plutonium dioxides (PuO〈sub〉2〈/sub〉), including their formation, migration, and effects on oxygen (O) self-diffusion, are systematically studied using atomic simulations in this work. Coulomb interactions among charged defects and ions are found play a key factor in determining the energetics and structures of defects in PuO〈sub〉2〈/sub〉, where point defects are energetically prefer to binding together forming defect pairs and less relaxation volumes are introduced by these bound point defects. Further calculations of defect migration properties reveal different migration mechanisms of O and Pu point defects, where O point defects prefer to migrate along [100] direction with an energy barrier of 0.31 eV by the vacancy mechanism, while Pu point defects prefer to migrate along [100] direction with an energy barrier of 2.57 eV by the interstitial mechanism. This confirms the cation sublattice is much more stable than the anion sublattice in the fluorite structure PuO〈sub〉2〈/sub〉. Given the dominance of O defects in PuO〈sub〉2〈/sub〉, their effects on O self-diffusion are investigated. Both vacancies and interstitials could significantly enhance O self-diffusivity. Under the Meyer-Neldel rule, activation energy and pre-exponential factors of O diffusion show a dependence of point defect concentrations. Specifically, internal stresses introduced by defects are found responsible for the different dependency of the activation energy on the vacancy and interstitial concentrations respectively. Finally, an empirical equation is derived to connect the point defect concentration, i.e. O/Pu ratio, and activation energy of O self-diffusion in PuO〈sub〉2〈/sub〉.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): M.H.A. Piro, M. Poschmann, P. Bajpai〈/p〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Guilin Wei, Bingsheng Li, Zhentao Zhang, Shunzhang Chen, Xiaoyan Shu, Xiao Wang, Yi Liu, Yi Xie, Dadong Shao, Xirui Lu〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉To immobilize the radioactive iodine isotopes, a series of borosilver silica gel glasses with iodine loaded in silver-coated silica gel ranging from 20 to 30 〈em〉wt〈/em〉% (I〈sup〉−〈/sup〉 content in each sample was 8 〈em〉wt〈/em〉%) had been successfully immobilized by muffle furnace from 450 °C to 550 °C for 6 h. With the increase of I〈sup〉−〈/sup〉 in silver-coated silica gel, optimum sintering temperatures corresponding to the maximum amorphous fraction in each system decreased from 550 °C (0.98) to 490 °C (0.68) and finally to 460 °C (0.66). The IR spectrums results show that the vitrified matrix is mainly consisted of [BO〈sub〉3〈/sub〉] and [SiO〈sub〉4〈/sub〉]. The SEM-EDS patterns show that iodine is uniformly distributed in the sample loaded with 20 〈em〉wt〈/em〉% of iodine in silver-coated silica gel. In addition, the specimen shows lower porosity with higher amorphous fraction. Our findings demonstrate the potential of boron for the immobilization of silver-coated silica gel with different radioactive iodine loadings.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Guanze He, Junliang Liu, Kexue Li, Jing Hu, Anamul Haq Mir, Sergio Lozano-Perez, Chris Grovenor〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The stability of the β-Nb Second Phase Particles (SPPs) in two types of Zr–Nb alloys (recrystallised Zr-1.0Nb and Zr-2.5Nb) was studied by in-situ heavy ion irradiation in a transmission electron microscope (TEM), combined with ex-situ analysis by energy dispersive x-ray spectroscopy (EDX). TEM thin foils were irradiated by 1 MeV Kr〈sup〉+〈/sup〉 ions at four different temperatures from 50 K to 873 K, and by 350 keV Kr〈sup〉+〈/sup〉 ions at different doses up to 39dpa. The change in size of individual β-Nb SPPs has been measured quantitatively, and the degradation mechanisms under irradiation at different temperatures discussed. It has been shown that the Nb redistribution between the SPPs and the Zr matrix is governed both by radiation induced mixing and local diffusion in the surrounding Zr matrix. Under the radiation conditions reported in this study, the β-Nb SPPs have shown remarkably stability against irradiation, and the extent of Nb redistribution between the SPPs and Zr matrix is very limited under all experimental conditions.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Hwasung Yeom, Benjamin Maier, Greg Johnson, Tyler Dabney, Mia Lenling, Kumar Sridharan〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Oxidation kinetics and microstructural evolution of cold sprayed Cr coatings on Zircaloy-4 at 1130–1310 °C in flowing steam at atmospheric pressure have been studied. The study is aimed at understanding of the response of Cr coated Zr-alloy under a steam environment and high temperatures pertinent to design basis accidents (DBAs) and beyond design basis accidents (BDBAs) in light water reactors (LWRs). Surface morphology, microstructure, and phases of post-oxidation test samples were characterized using Scanning Electron Microscopy (SEM), x-ray diffraction (XRD), and Scanning Transmission Electron Microscopy (STEM). Growth kinetics of the Cr-oxide scale and interdiffusion layers between the Cr coating and the Zr-alloy substrate were quantified from cross-sectional SEM images. Cross-sectional analysis showed that the Cr coatings offered a 50 times reduction in oxidation rate over bare Zircaloy-4 in a 1310 °C steam environment. Oxidation kinetics at 1130 °C followed parabolic law (i.e., n ∼ 0.5 in power law kinetics) but at 1230 °C and 1310 °C the value of n was suppressed to below 0.5 possibly due to the volatilization of Cr species at the two highest temperatures. Interdiffusion at coating/substrate interface resulted in formation of a brittle Cr〈sub〉2〈/sub〉Zr or Zr(Fe, Cr)〈sub〉2〈/sub〉 intermetallic compound layer on the order of micrometer in thickness and scattered Cr-rich precipitates were observed well below the interface within the Zr-alloy substrate after cooling. These experimental results could provide data to LWR system simulation codes for better estimation of coping time in the event of accidents.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Divya Singh, Avinash Parashar, A. Kedharnath, Rajeev Kapoor, Apu Sarkar〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉In this article, molecular dynamics based simulations were carried out to study the fracture behaviour of bi-crystalline zirconium (Zr). Atomistic simulations were performed to study the effect of grain boundary configuration on the crack tip behaviour subjected to opening mode of loading. Separate set of simulations were carried out to study the influence of crack orientation on the failure morphology, and strength of bi-crystalline Zr. Disclination shielding of the crack tip stresses induced by grain boundaries, significantly affects the fracture behaviour of bi-crystalline Zr. Opening stresses at the crack tip varied as a function of distance between crack tip and grain boundary plane. Crack propagation and blunting of tip was also predicted as the function of distance from the grain boundary plane. Effect of irradiation induced point defects have also been studied with respect to spatial distribution and number of defects.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Yingtao Wang, Huadong Fu, Jian Wang, Hongtao Zhang, Weidong Li, Jianxin Xie〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉With the goals of exploring the feasibility of replacing Zr with Fe and P in Cu–Cr–Zr alloys and developing Cu–Cr–Fe–P alloys with excellent mechanical and electrical properties, the effects of Fe and P additions on the microstructures and properties of the as-cast, cold-rolled and aged Cu-0.8Cr alloys are investigated. The results indicate that adding Fe and P into the Cu-0.8Cr alloy can promote the precipitation of a large amount of fine, uniformly dispersed Cr and Cr〈sub〉3〈/sub〉P particles. From Cu-0.8Cr to Cu-0.8Cr-0.17Fe-0.048P, the area fraction of precipitates increases from 3.1% to 6.3% while their average diameter reduces from 1.6 μm to 1.4 μm. When subjected to 95% cold rolling plus 450 °C aging for 1 h, the Cu-0.8Cr-0.17Fe-0.048P alloy exhibits a Vickers hardness of 140.4 HV and an electrical conductivity of 73.2%IACS, higher than 96.7 HV and 58.4%IACS in the Cu-0.8Cr alloy. Among three different processing routes investigated, the one with initial 50% cold rolling → aging at 500 °C for 1 h → final 95% cold rolling allows the Cu-0.8Cr-0.17Fe-0.048P alloy to gain most in both its tensile strength (569 MPa) and electrical conductivity (64%IACS). The findings in this work provide insights to composition and processing designs for new Cu–Cr alloys with high strength and electrical conductivity.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Lei Li, Shuoxue Jin, Peng Zhang, Dandan Wang, Xingzhong Cao, Liping Guo, Qiu Xu, Jun Li, Tongmin Zhang, Liben Li, Baoyi Wang〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉In order to investigate the effect of interaction between H and He ions on micro-defects in Fe9Cr alloy, well-annealed Fe9Cr alloy were pre-damaged firstly with 1 MeV Ni ions to a dose of 3.4 × 10〈sup〉13〈/sup〉/cm〈sup〉2〈/sup〉, and then irradiated with single (H or He) ion beam and sequential (H/He or He/H) ions beam. The irradiation were performed with 50 keV H ions, 80 keV He ions with doses of 1 × 10〈sup〉16〈/sup〉/cm〈sup〉2〈/sup〉 and 2 × 10〈sup〉15〈/sup〉/cm〈sup〉2〈/sup〉 at 723 K, respectively. Results of slow positron beam show that the S parameter decreased obviously after single (H or He) ions beam irradiation compared to the pre-damaged specimen, which indicate H or He (H〈sub〉n〈/sub〉 or He〈sub〉n〈/sub〉) clusters deposit at the damage area caused by Ni ion irradiation and they occupy vacancy sites (V〈sub〉m〈/sub〉), resulting in the formation of numerous H〈sub〉n〈/sub〉V〈sub〉m〈/sub〉 or He〈sub〉n〈/sub〉V〈sub〉m〈/sub〉 clusters. The S parameter induced by He/H ions sequential irradiation didn't change significantly compared to that by single He ion irradiation. However, obvious increment of the S parameter induced by H/He ion sequential irradiation is higher than that induced by singe H ion irradiation. The binding energy of H〈sub〉n〈/sub〉V〈sub〉m〈/sub〉He is larger than He〈sub〉n〈/sub〉V〈sub〉m〈/sub〉H, and it is difficult for subsequent H ions irradiation to decompose the He〈sub〉n〈/sub〉V〈sub〉m〈/sub〉 complexes.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): I. Ipatova, R.W. Harrison, S.E. Donnelly, M.J.D. Rushton, S.C. Middleburgh, E. Jimenez-Melero〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉We have probed void evolution in polycrystalline W and W-5wt.%Ta material at 800 and 1000 °C, by transmission electron microscopy during in-situ irradiation with a 40 keV proton beam. The presence of radiation-induced dislocation loops was not observed prior to void formation at those elevated temperatures. The damaged W microstructure was characterised by the presence of a population of randomly distributed voids, whose number density reduces when the irradiation temperature increases. Soft impingement of voids becomes noticeable at damage levels ≥0.2 dpa. In contrast, the excess of free vacancies in the W-5wt.%Ta material irradiated at 800 °C only leads to the formation of visible voids in this TEM study (≥2 nm) after post-irradiation annealing of the sample at 1000 °C. Solute Ta atoms also cause a significant increase in the number density of voids when comparing the microstructure of both materials irradiated at 1000 °C, and a gradual progression towards saturation in average void size at ≥0.2 dpa. Moreover, we have detected a progressive transition from a spherical to a faceted shape in a number of voids present in both materials at damage levels ≥0.3 dpa.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311519304908-fx1.jpg" width="466" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Da Chen, Shijun Zhao, Jianrong Sun, Pengfei Tai, Yanbin Sheng, Yilu Zhao, Guma Yeli, Weitong Lin, Shaofei Liu, Wu Kai, Ji-Jung Kai〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉For the structure materials applied in advanced nuclear energy system, helium bubble formation is always a big concern which will severely degrade the performance of materials around or above the half-melting temperature regime (∼0.5 〈em〉T〈/em〉〈sub〉〈em〉m〈/em〉〈/sub〉). To explore the He bubble formation resistance in the FeCoNiCr alloy, that is a novel face-centered cubic (fcc) high-entropy alloy (HEA) showing excellent radiation damage tolerance, we conducted a series of 2 MeV He ions irradiation experiments on them at three different temperatures (0.46, 0.51 and 0.57 〈em〉T〈/em〉〈sub〉〈em〉m〈/em〉〈/sub〉). For reference purpose, a model 〈em〉fcc〈/em〉 metallic system of pure Ni was irradiated simultaneously. Through transmission electron microscopy (TEM), He bubble formation in the irradiated samples was systematically investigated. The results show that in any designated temperature, He bubbles have a smaller size, higher number density, and denser distribution in the HEA when comparing to that of pure Ni. The volume fraction of He bubbles is also less in the HEA, suggesting a suppressed bubble evolution. For the underlying mechanism of the He bubble formation resistance of HEA, we suggest that the featured energy barriers for point defects migration in the HEA will promote the recombination of defects and somewhat reduce the vacancy concentration during irradiation. Such unique effect could suppress the He diffusion through vacancy mechanism, it will finally influence the evolution of He bubbles in the HEA.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Cheol Min Lee, Gyeonghun Kim, Dong-Seong Sohn, Young-Soo Han, Yong-Kyoon Mok〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The oxidation of zirconium alloys is largely influenced by cracks in the oxide layers. It was reported that cracks formed in ZrO〈sub〉2〈/sub〉/Al〈sub〉2〈/sub〉O〈sub〉3〈/sub〉 can be healed by the phase transformation of ZrO〈sub〉2〈/sub〉. Here, we conducted several experiments to determine whether this also holds true for zirconium alloys. The experimental results showed that the cracks at the oxide-metal interface were healed after oxidation at 1200 °C, but not healed after oxidation at 1000 and 1100 °C. It appears that this crack healing behavior is related to the phase transformation of ZrO〈sub〉2〈/sub〉.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311519307767-fx1.jpg" width="500" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): D. Frazer, D. Jadernas, N. Bolender, J. Madden, J. Giglio, P. Hosemann〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Proliferation concerns are the driving force to reduce the enrichment of research reactor fuel from highly enriched fuel to lower levels of enriched fuel. One promising fuel alloy that would enable a reduction in enrichment without greatly diminishing fuel performance is Uranium with 10 wt % Molybdenum (U-10 wt-% Mo). While there is a large amount of data available on the microstructure and thermal properties of this new fuel type currently there is insufficient data available on the mechanical properties. Small scale mechanical testing techniques can be used to evaluate the mechanical properties of the as-fabricated U-10 wt% Mo components at their intended operating temperature which should improve the accuracy of the input parameters employed by modelers. In this study the mechanical properties of a U-10 wt% Mo fuel meat sample taken from an as-fabricated fuel plate for a new fuel design were evaluated using in-situ scanning electron microscopy microcantilever testing under varying temperatures between room temperature and 200 °C.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): P. Souček, K. Uruga, T. Murakami, A. Rodrigues, S. Van Winckel, M. Iizuka, J.-P. Glatz〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉An electrorefining process for homogeneous group-selective recovery of actinides from metallic nuclear fuel in molten LiCl–KCl has been investigated. The present study follows up on the previously achieved results on recovery of actinides from both non-irradiated and irradiated test metallic fuels. In the current work, METAPHIX-2 fuel initially composed of U〈sub〉71〈/sub〉–Pu〈sub〉19〈/sub〉-Zr〈sub〉10〈/sub〉 alloy irradiated to ∼7 at.% was processed. The experiments were focused on evaluation of selectivity of actinides over lanthanides during the electrorefining process in an electrolyte containing different concentrations of dissolved lanthanides up to 6.5 wt%, simulating the later and final stages of the process. In addition, a comparison of use of the solid reactive aluminium and solid inert cathodes for homogeneous recovery of all actinides was studied, as well as the effect of zirconium co-dissolution from the fuel on the process efficiency and on the structure of the deposits. The reactive aluminium electrode was proven suitable for homogeneous recovery of all actinides, while at the given conditions only uranium could be deposited on the inert cathodes. Very high group-selectivity of the process for actinides was demonstrated, even at high concentration of lanthanides in the electrolyte.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Li Jiang, Pengyuan Xiu, Yan Yan, Chenyang Lu, Minjiang Huang, Tong Liu, Chao Ye, Haiping Sun, Rui Shu, Lumin Wang〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Recently, the development of accident tolerant fuel is a critical issue in light water reactors. Cr coatings display a promising application in protecting Zr alloys from oxidation and maintaining the fuel performance. In this work, an attempt has been made to systematically investigate the dose-dependent and thickness-related irradiation responses of Cr coatings on the Zr alloy. Three sets of Cr coatings on a Zr alloy with thickness of 5, 10 and 12 μm have been irradiated at 400 °C with 6 MeV Au ions to average doses of 10, 25 and 50 dpa. Pre- and post-irradiation microstructures were characterized with transmission electron microscopy and radiation hardening was evaluated by nano-indentation. An obvious thickness dependent columnar grain size is observed before radiation with smaller grains in thinner coatings. Although the sizes of dislocation loops increase with irradiation dose in all samples, the evolution of irradiation defects is delayed in thinner coatings that can be attributed to the smaller columnar grains and larger lattice distortion.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311519309304-fx1.jpg" width="341" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): J.A. Sawicki〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The properties of Fe additions in Zr-2.5Nb pressure tube alloy are investigated after plastic deformations, heat treatments below its monotectoid temperature and severe radiation damage by neutrons and ions. Mössbauer spectra showed that in the β-quenched and hot-extruded Zr-2.5Nb about 80% of Fe atoms is located in βZr and the reminder in a metastable ω-phase formed within βZr filaments retained in αZr grain boundaries. The fraction of Fe in ω-phase increased to ∼40% after heat treatment at 400 °C for 24 h in vacuum, and also after intense cold rolling, service in the reactor core and ion implantation. In samples with β-phase fully transformed at ∼560–580 °C to βNb, Fe in ω-phase didn't appear after any cold work, while in untransformed samples the fraction of Fe in ω-phase clearly increased with the extent of cold-rolling. A large 〈em〉s〈/em〉-electron density observed at 〈sup〉57〈/sup〉Fe nuclei in ω-phase suggests that iron atoms in this phase are mostly located in low-volume B-sites, with hybridized 〈em〉sd〈/em〉〈sup〉〈em〉2〈/em〉〈/sup〉 trigonal, partially covalent orbitals.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Thomas S. Neill, Katherine Morris, Carolyn I. Pearce, Liam Abrahamsen-Mills, Libor Kovarik, Simon Kellet, Bruce Rigby, Tonya Vitova, Bianca Schacherl, Samuel Shaw〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉U(IV) mobility can be significantly enhanced by colloids in both engineered and natural environments. This is particularly relevant in decommissioning and clean-up of nuclear facilities, such as legacy fuel ponds and silos at the Sellafield site, UK, and in long-term radioactive waste geodisposal. In this study, the product of metallic uranium (U) corrosion under anaerobic, alkaline conditions was characterised, and the interaction of this product with silicate solutions was investigated. The U metal corrosion product consisted of crystalline UO〈sub〉2〈/sub〉 nanoparticles (5–10 nm) that aggregated to form clusters larger than 20 nm. Sequential ultrafiltration indicated that a small fraction of the U metal corrosion product was colloidal. When the uranium corrosion product was reacted with silicate solutions under anaerobic conditions, ultrafiltration indicated a stable colloidal uranium fraction was formed. Extended X-ray absorption fine structure (EXAFS) spectroscopy and high resolution TEM confirmed that the majority of U was still present as UO〈sub〉2〈/sub〉 after several months of exposure to silicate solutions, but an amorphous silica coating was present on the UO〈sub〉2〈/sub〉 surface. This silica coating is believed to be responsible for formation of the UO〈sub〉2〈/sub〉 colloid fraction. Atomic-resolution scanning TEM (STEM) indicated some migration of U into the silica-coating of the UO〈sub〉2〈/sub〉 particles as non-crystalline U(IV)-silicate, suggesting alteration of UO〈sub〉2〈/sub〉 at the UO〈sub〉2〈/sub〉-silica interface had occurred. This alteration at the UO〈sub〉2〈/sub〉-silica interface is a potential pathway to the formation of U-silicates (e.g. coffinite, USiO〈sub〉4〈/sub〉).〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311519306464-fx1.jpg" width="491" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Sanjit Kumar Parida, S. Nagaraj, M. Venkatesh, R. Sudha, Kuntla Reddy Sekhar, R. Ramaseshan, Rajesh Ganesan〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Interaction of steel specimens such as SS 316 LN, D9, 9Cr–1Mo, oxide dispersed steel (ODS) and pure iron with liquid lithium has been studied at 823 and 973 K (550 and 700 °C) for various durations from 250 to 1000 h at 250 h intervals. The relevance of this study is to identify a suitable container material for liquid lithium which has been proposed as a liquid poison to control the reactivity in nuclear reactors. The metallic specimens were characterised before and after exposure to liquid lithium by X-ray diffraction (XRD), scanning electron microscopy (SEM) and energy dispersive X-ray analysis (EDX) and mechanical testing. The results indicate that except for pure iron, lithium interacts with all steel samples in varying degree. The depth of interaction varies from 20 to 60 μm depending on nature of specimen, temperature and time of exposure.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Justin Hesterberg, Zhijie Jiao, Gary S. Was〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Post-irradiation annealing (PIA) was conducted on a 304L stainless steel irradiated to 5.9 dpa in the Barsebäck-1 BWR reactor, to investigate its effect on the mitigation of irradiation-assisted stress corrosion cracking (IASCC) susceptibility. IASCC susceptibility was measured for the as-irradiated and four PIA conditions (500 °C: 1 h and 550 °C: 1, 5, and 20 h) via interrupted constant extension rate tensile and four-point bend experiments under simulated BWR-NWC conditions. The annealing treatments were observed to progressively reduce IASCC susceptibility, as measured by the final intergranular fracture fraction (tensile) and crack length per unit area (four-point bend), with full removal of IASCC susceptibility being observed following annealing at 550 °C: 1 h for tensile specimens and 500 °C: 1h for four-point bend specimens. Among the microstructure and mechanical property parameters measured as a function of PIA, the average dislocation channel spacing was observed to decrease by ∼25% and ∼40% from the as-irradiated condition after annealing at 500 °C: 1 h and 550 °C: 1 h, respectively. The mitigation of IASCC susceptibility correlated well with the decrease in the average dislocation channel spacing and is consistent with a process in which crack initiation is controlled in part by the high tensile stress at dislocation channel-grain boundary intersections.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Karen E. Wright, Jason M. Harp, Luca Capriotti〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉 〈p〉Electron microprobe examinations were performed to characterize the chemical features of a full cross section of irradiated nuclear fuel from the FUTURIX-FTA experiment. This experiment investigated the nuclear fuel performance of a candidate fuel alloy intended for the transmutation of long-lived minor actinides in a fast neutron spectrum. The irradiated fuel, designated FUTURIX-FTA DOE1, was composed of 34.1U-28.3Pu-3.8Am-2.1Np-31.7Zr (where the preceding numbers indicate concentrations in weight %). The fuel was irradiated in the Phénix sodium fast reactor in France to a measured burnup of 9.5% fissions per initial heavy metal atom (FIMA), and experienced a peak cladding temperature of 550 °C.〈/p〉 〈p〉Microprobe analysis showed elemental redistribution of Zr and U where Zr has increased in concentration in the fuel center from an initially fabricated content of 31.7 wt % to 41.5 wt%, and U decreased from 34.1 wt% to 24.8 wt%. From the center of the fuel extending out radially approximately 1 mm, the fuel represented dominantly a single phase. Beyond this region to the fuel periphery, the fuel separated into two major phases, descibed by their composition as a (U, Np, Pu) Zr〈sub〉2〈/sub〉-like phase and a high uranium content-low zirconium content phase. From the outer radius of the fuel extending approximately 1.7 mm radially into the fuel, americium, lanthanide elements, and actinide elements precipitated in a phase whose chemical analysis resembles Nd〈sub〉7〈/sub〉(Pd, Rh)〈sub〉3〈/sub〉. In addition, americium occurred as a dissolved species in the major fuel phases. Sm and Am penetrated up to 15 μm into the cladding along presumed grain boundaries, while major cladding elements Fe, Ni, and Cr penetrated at least 30 μm into the fuel. No phase formation between cladding elements and fuel elements was observed as the result of cladding element diffusion into the fuel.〈/p〉 〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Sungjune Sohn, Gwan Yoon Jeong, Seongjin Jeong, Jungho Hur, Heejae Ju, Yong-Hoon Shin, Jaeyeong Park, Il Soon Hwang〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉This study presents the spatial distribution of CeBi〈sub〉2〈/sub〉 in Bi-Ce alloys formed in various conditions to investigate density-based separation using liquid Bi between the intermetallic compound of actinides and lanthanides in used molten salt from pyroprocessing. It was experimentally identified that CeBi〈sub〉2〈/sub〉 clearly formed and floated at the top of Bi-Ce alloy. Four Bi-Ce alloys were metallurgically prepared to investigate the feasibility of floating intermetallic compounds with the consideration of Ce concentrations in the liquid Bi alloy and cooling rates. Cyclic voltammetry (CV) of CeCl〈sub〉3〈/sub〉 in LiCl-KCl was conducted to examine the electrochemical behavior of Ce ion on liquid Bi pool electrode at 500 °C. The Bi-Ce alloy formed by the galvanostatic electrolysis of LiCl-KCl-CeCl〈sub〉3〈/sub〉 on Bi cathode at 500 °C. The applied cathodic current was determined to be 20 mA/cm〈sup〉2〈/sup〉 based on the CV results. The spatial distribution of intermetallic compounds was obtained by scanning electron microscope images which focused on the vertically entire cross sections of all Bi-Ce alloys. The intermetallic phase was characterized by both energy dispersive spectroscopy and X-ray diffractometry. From the experimental results, we suggest the feasibility of the density-based separation process and its flowsheet for the application to pyroprocessing technology.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S002231151930409X-fx1.jpg" width="500" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Mohammed Imtyazuddin, Anamul H. Mir, Matheus A. Tunes, Vladimir M. Vishnyakov〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Cr〈sub〉2〈/sub〉AlC MAX phases were deposited using magnetron sputtering. The synthesis was performed via layer-by-layer deposition from elemental targets onto Si wafer and polished Inconel® 718 superalloy substrates at 650 K and 853 K. Transmission Electron Microscopy (TEM) characterisation showed that the thin films had a thickness of about 0.8 and 1.2 μm for Si and Inconel® substrates, respectively, and a MAX phase crystalline structure. Depositions onto Inconel substrate was performed in order to measure film mechanical properties. The films have hardness at around 15 GPa, reduced Young's modulus at around 260 GPa, do not delaminate and showed characteristic ductile behaviour during nanoscratching. Ion irradiations with 〈em〉in situ〈/em〉 TEM were performed with 320 keV Xe〈sup〉+〈/sup〉 ions up to fluence 1 × 10〈sup〉16〈/sup〉 ions·cm〈sup〉−2〈/sup〉 at 300 K and 623 K. At 300 K the Cr〈sub〉2〈/sub〉AlC started to amorphise at around 0.3 dpa. At displacement levels above 3.3 dpa all crystalline structure was almost completely lost. Conversely, irradiations at 623 K showed no recordable amorphisation up to 90 dpa. It is discussed that the presence of many grain boundaries and low defect recombination energy barriers are responsible for high radiation hardness of Cr〈sub〉2〈/sub〉AlC MAX phase at 623 K. The thin film Cr〈sub〉2〈/sub〉AlC MAX phases have mechanical and radiation stability which makes them a candidate for fuel rod coating as Accident Tolerant Fuels (ATF) material for the next generation of nuclear reactors.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 526〈/p〉 〈p〉Author(s): Pengyan Mao, Jingping Cui, Yangchun Chen, Jianhang Qiu, Qun Jin, Jixiang Qiao, Yang Zhao, Kan Cui, Ning Gao, Kaiping Tai〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉It has long been recognized that grain boundary (GB) is an effective sink to trap irradiation-induced defects. Introducing large densities of GBs becomes an effective strategy to enhance irradiation resistance. However, the nano-grained materials have poor stability under the extreme irradiation, and the effect of nano-GB characters on the irradiation sink strength is largely unknown. Here, the sink strength of nano-GB is quantitatively investigated in the nano-grained Cu and dilute Cu–W alloys with the average grain size ranging from ∼20 to ∼50 nm by ∼300 keV He ions irradiation at room temperature and 673 K. The irradiation induced void volume ratio, size and distribution are confirmed to strongly depend on the grain size and irradiation temperature. The nano-GBs with different characters, such as the misorientation and GB plane, have similar relative energy. The nano-GB sink strength is independent on the GB characters, deriving from the highly curved nano-GB plane having excess volume and energy. Combining with molecular dynamics simulations, we can conclude that nano-grain size is the most vital factor for the sink strength with respect to the GB characters. With the increase of temperature and the decrease of grain size, the stable nano-GBs exhibit a behavior of “ideal” defects sink due to their high volume ratio and the increased point defect recombination probability. Our work provides a fundamental understanding of the nano-GB sink efficiency and offer a guidance for designing nano-grained structural materials with optimum anti-irradiation performance for future fusion reactors.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311519303824-fx1.jpg" width="500" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): Matic Pečovnik, Sabina Markelj, Anže Založnik, Thomas Schwarz-Selinger〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The influence of grain size on deuterium transport and retention in tungsten was studied. For this purpose an experiment was carried out on three polycrystalline tungsten samples with different grain sizes and a single crystal sample with surface orientation 〈100〉. In order to increase deuterium retention and hence the sensitivity for detection, samples were first damaged by high energy W ions. After damaging, the samples were exposed to a flux of deuterium atoms at 600 K for 70 h. During the exposure the depth profile of the retained deuterium was measured by Nuclear Reaction Analysis using a 〈sup〉3〈/sup〉He ion beam. After the exposure the samples were also analysed by Thermal Desorption Spectroscopy. A clear difference in the time dependence of deuterium uptake was noticed between different samples. The experimental results were modeled using a rate-equation model. The influence of different grain size was modeled by changing the effective height of the potential barrier for deuterium atoms to enter into the bulk. We managed to successfully describe the transport of deuterium into the bulk of tungsten by reducing the potential barrier for samples with smaller grain sizes while the barrier for the sample with larger grain size was close to the value for the damaged single crystal sample.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311518307256-fx1.jpg" width="256" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): P.A. Burr, E. Kardoulaki, R. Holmes, S.C. Middleburgh〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The stability, diffusivity and clustering behaviour of defects in uranium diboride (UB〈sub〉2〈/sub〉) was investigated in light of the potential application as a burnable absorber in nuclear fuel. UB〈sub〉2〈/sub〉 was found to accommodate limited deviations from stoichiometry, which should be a consideration when manufacturing and operating the material. Self-diffusivity of both U and B was found to be sluggish (10〈sup〉−14〈/sup〉 cm〈sup〉2〈/sup〉/s for B and 10〈sup〉−19〈/sup〉 cm〈sup〉2〈/sup〉/s for U at 2000K) and highly anisotropic, with migration along the basal planes being orders of magnitude faster than 〈em〉c〈/em〉-axis migration. The anisotropy of defect migration (both interstitials and vacancies) is predicted to hinder recombination of defects produced by collision cascades, thus limiting the radiation tolerance of the material. Boron and uranium vacancies exhibit a drive to cluster. Boron vacancies in particular, which are mobile on basal planes, are predicted to cluster into strongly bound di-vacancy, which in turn are less mobile. These are then predicted to grow into larger two-dimensional vacancy clusters on the B plane, leading to anisotropic swelling. We provide an analytical expression to predict the stability of these clusters based on purely geometrical considerations. Finally, the accommodation of Li, He and Xe onto vacancy clusters was considered. Li appears to stabilise the structure upon U depletion, while the retention of He and Xe appears to rise with increasing B depletion, through the formation of vacancy clusters.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): Fusheng Li, Shilei Li, Gang Liu, Xudong Chen, Yanli Wang〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉 〈p〉Zircaloy-4 with 200 wppm hydrogen was irradiated with 3 MeV Fe〈sup〉11+〈/sup〉 ions and the microstructural evolution of δ-hydride precipitation under ion irradiation was investigated. TEM observation of unirradiated sample shows the needle-like δ-hydride and the orientation relationship between α-Zr matrix and δ-hydride is (0002)〈sub〉α〈/sub〉//(1 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.gif" overflow="scroll"〉〈mrow〉〈mover accent="true"〉〈mn〉1〈/mn〉〈mo〉¯〈/mo〉〈/mover〉〈/mrow〉〈/math〉 1)〈sub〉δ〈/sub〉, [2 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si2.gif" overflow="scroll"〉〈mrow〉〈mrow〉〈mover accent="true"〉〈mn〉1〈/mn〉〈mo〉¯〈/mo〉〈/mover〉〈/mrow〉〈mrow〉〈mover accent="true"〉〈mn〉1〈/mn〉〈mo〉¯〈/mo〉〈/mover〉〈/mrow〉〈/mrow〉〈/math〉 0]〈sub〉α〈/sub〉//[011]〈sub〉δ〈/sub〉. After ion irradiation, both 〈a〉 and 〈c〉-type dislocation loops were observed in α-Zr matrix.〈/p〉 〈p〉Some circular structures with obviously different contrast from the α-Zr matrix were observed in the irradiated Zircaloy-4-200 wppm H samples. The circular structures were identified as δ-hydrides with an fcc structure by indexing the FFT patterns of HRTEM images. There is an orientation relationship of (01 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si2.gif" overflow="scroll"〉〈mrow〉〈mrow〉〈mover accent="true"〉〈mn〉1〈/mn〉〈mo〉¯〈/mo〉〈/mover〉〈/mrow〉〈mrow〉〈mover accent="true"〉〈mn〉1〈/mn〉〈mo〉¯〈/mo〉〈/mover〉〈/mrow〉〈/mrow〉〈/math〉 )〈sub〉α〈/sub〉//(200)〈sub〉δ〈/sub〉, [01 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.gif" overflow="scroll"〉〈mrow〉〈mover accent="true"〉〈mn〉1〈/mn〉〈mo〉¯〈/mo〉〈/mover〉〈/mrow〉〈/math〉 2]〈sub〉α〈/sub〉//[011]〈sub〉δ〈/sub〉 between the circular δ-hydride and α-Zr. The formation of circular δ-hydride is related with ion irradiation and the formation process is explained by the hydrogen trapping mechanism.〈/p〉 〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): Miao Song, Mi Wang, Xiaoyuan Lou, Raul B. Rebak, Gary S. Was〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The irradiation-induced microstructure and irradiation-assisted stress corrosion cracking (IASCC) behavior of additively manufactured (AM) 316L stainless steels produced by laser powder bed fusion were evaluated for the first time. Irradiation-induced dislocation loops, voids, and γ′ precipitates were observed in all processing conditions following 2.5 dpa at 360 °C. The cell structure and dense dislocation walls in the stress-relieved AM materials recovered and showed signs of recrystallization following irradiation. Anisotropy in both tensile property and IASCC susceptibility were observed in the stress-relieved AM 316L stainless steel due to the printing texture. The hot-isotropic pressed AM 316L had better radiation tolerance and lower IASCC susceptibility than the stress-relieved AM 316L and conventionally forged 316L. Therefore, post-printing hot-isotropic pressing (HIP) is recommended for enhancing radiation tolerance and IASCC performance in nuclear applications as it eliminates the anisotropic mechanical behavior and IASCC susceptibility associated with the printing texture.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): M.J. Rahman, B. Szpunar, J.A. Szpunar〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The thermomechanical properties of mixed actinide oxide and its dependence on actinide composition are of significant interest for the safety in reactor design and operation. In this study, the elastic properties of (U〈sub〉x〈/sub〉,Th〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉, (U〈sub〉x〈/sub〉,Pu〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉 and (Pu〈sub〉x〈/sub〉,Th〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉 solid solutions for 50-50 cation composition and thermal conductivity of (Pu〈sub〉x〈/sub〉,Th〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉 system, for x = 0.3 and 0.5, have been investigated in the temperature range from 300 to 1500 K using Molecular dynamics (MD) simulations. Compared to pure oxides, reduced thermal conductivity is reported for (Pu〈sub〉x〈/sub〉,Th〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉 mixture due to non-uniform cation sublattice. The degree of reduction in conductivity is the largest in (Pu〈sub〉x〈/sub〉,Th〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉 than that of previously published (U〈sub〉x〈/sub〉,Th〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉 and (U〈sub〉x〈/sub〉,Pu〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉. This is attributed to the largest mismatch in lattice parameter between PuO〈sub〉2〈/sub〉 and ThO〈sub〉2〈/sub〉. In general, the elastic properties, for all pure and mixed oxide fuels, linearly decrease with the increase in temperature, and this is commonly observed also in other studies. The rate of change in elastic modulus (〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.gif" overflow="scroll"〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈/math〉) with temperature follows the trend, for pure oxides: 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si2.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉P〈/mi〉〈mi〉u〈/mi〉〈mi〉O〈/mi〉〈mn〉2〈/mn〉〈/mrow〉〈/msub〉〈mo〉〉〈/mo〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉U〈/mi〉〈mi〉O〈/mi〉〈mn〉2〈/mn〉〈/mrow〉〈/msub〉〈mo〉〉〈/mo〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉T〈/mi〉〈mi〉h〈/mi〉〈mi〉O〈/mi〉〈mn〉2〈/mn〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉 and for mixed oxides: 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si3.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉U〈/mi〉〈mi〉P〈/mi〉〈mi〉u〈/mi〉〈mi〉O〈/mi〉〈mn〉2〈/mn〉〈/mrow〉〈/msub〉〈mo〉〉〈/mo〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉U〈/mi〉〈mi〉T〈/mi〉〈mi〉h〈/mi〉〈mi〉O〈/mi〉〈mn〉2〈/mn〉〈/mrow〉〈/msub〉〈mo〉〉〈/mo〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉P〈/mi〉〈mi〉u〈/mi〉〈mi〉T〈/mi〉〈mi〉h〈/mi〉〈mi〉O〈/mi〉〈mn〉2〈/mn〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉. In most cases, for (U〈sub〉x〈/sub〉,Th〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉 and (U〈sub〉x〈/sub〉,Pu〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉, the elastic modulus is similar to the linear interpolation between the corresponding pure oxides. However, for (Pu〈sub〉x〈/sub〉,Th〈sub〉1-x〈/sub〉)O〈sub〉2〈/sub〉, significant deviation from the linearity is observed and that is justified by differences in oxygen mobility in the mixed oxide fuels. In addition, for all oxides, except Pu〈sub〉0.5〈/sub〉Th〈sub〉0.5〈/sub〉O〈sub〉2〈/sub〉, our results predict 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si4.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉B〈/mi〉〈/mrow〉〈/msub〉〈mo〉〉〈/mo〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉Y〈/mi〉〈/mrow〉〈/msub〉〈mo〉〉〈/mo〉〈msub〉〈mrow〉〈mi〉ε〈/mi〉〈/mrow〉〈mrow〉〈mi〉G〈/mi〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉 where 〈em〉B〈/em〉, 〈em〉Y〈/em〉 and 〈em〉G〈/em〉 being the bulk elastic properties. A calculation of elastic modulus at 0 K for all oxides is presented.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): T. Narayana Murty, R.N. Singh, P. Ståhle〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Formation of partially constrained precipitates such as hydride blisters and oxide nodules have been reported on surfaces of Zr-alloy components of pressurised heavy water reactors and is associated with a large increase in volume. Such a change in volume imposes large stresses in the material surrounding the precipitate and may facilitate stable crack growth through delayed hydride cracking. In this work, the stress field of the partially constrained precipitates with different depth and aspect ratio has been computed using a finite element method. The computed stress field is used to predict the region in the matrix in which radial hydride is likely to form and fracture, by taking into consideration grain-size, texture and multi-axial state of stress. For a hypothetical crack just below the precipitate, stress intensity factors are estimated using material properties for both unirradiated and irradiated pressure tube materials. The results are compared with the threshold stress intensity factors required for crack growth due to delayed hydride cracking.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉In Fig. 1(a) and (b) plot of 〈em〉KJ〈/em〉 as a function of aspect ratio of oxide nodules for different depths using unirradiated and irradiated material properties is shown. Also depicted in this figure are the reported values of threshold stress intensity factors above which crack growth by Delayed Hydride Cracking is possible. It is evident form figures that in cases of unirradiated material, oxide nodules having aspect ratios less than two and depths more than 350 μm can initiate DHC whereas nodules having aspect ratios less than 2.5 and depths more than 200 μm can initiate DHC in irradiated material.〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311518308985-fx1.jpg" width="126" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): G. Singh, J. Gorton, D. Schappel, N.R. Brown, Y. Katoh, B.D. Wirth, K.A. Terrani〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composites are being considered as components in light water reactor cores to improve accident tolerance, including channel boxes and fuel cladding. In the nuclear reactor environment, core components like a channel box will be exposed to neutron and other radiation damage and temperature gradients. To ensure reliable and safe operation of a SiC-SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one year. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions that have been calculated using the neutronics and thermal-hydraulics codes Serpent and CTF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5 mm. The channel box bowing behavior is time dependent and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 15 December 2018〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 512〈/p〉 〈p〉Author(s): Jin-Wen Yang, Li An〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉We have investigated the structure, elastic anisotropy, lattice dynamics, ideal tensile and shear strengths of USi〈sub〉3〈/sub〉 and U〈sub〉3〈/sub〉Si using first-principles method of density functional theory (DFT). The present fully relaxed structural parameters, equilibrium volumes as well as the single crystal elastic constants agree well with the available experimental data and other theoretical results. Also, the polycrystalline properties such as bulk modulus, Young's modulus, and shear modulus, Poisson's ratio and brittle/ductile properties have been evaluated using Voigt-Reuss-Hill model. The elastic anisotropy of 〈em〉Fmmm〈/em〉 U〈sub〉3〈/sub〉Si has been characterized by different anisotropic factors. It is concluded that 〈em〉Pm-〈/em〉3m USi〈sub〉3〈/sub〉, 〈em〉Pm-〈/em〉3m U〈sub〉3〈/sub〉Si, and 〈em〉Fmmm〈/em〉 U〈sub〉3〈/sub〉Si should be stabilized mechanically, and 〈em〉Pm-〈/em〉3m USi〈sub〉3〈/sub〉 should be stabilized mechanically up to 80 GPa. The dynamical properties of U〈sub〉3〈/sub〉Si and USi〈sub〉3〈/sub〉 have been investigated using density functional perturbation theory (DFPT) method. The present calculations mean that both 〈em〉Pm-〈/em〉3m USi〈sub〉3〈/sub〉 and 〈em〉Fmmm〈/em〉 U〈sub〉3〈/sub〉Si are dynamically stable, and 〈em〉Pm-〈/em〉3m USi〈sub〉3〈/sub〉 should be stabilized dynamically up to 80 GPa. However, 〈em〉Pm-〈/em〉3m U〈sub〉3〈/sub〉Si is not dynamically stable. Additionally, the ideal tensile strengths along typical crystallographic directions, and the ideal shear strengths in the (010)[101] slip system of 〈em〉Pm-〈/em〉3m USi〈sub〉3〈/sub〉 and 〈em〉Fmmm〈/em〉 U〈sub〉3〈/sub〉Si have been explored using first principles total energy method. These calculated results provide useful information for the applications of uranium silicides as potential nuclear fuels.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2018〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 511〈/p〉 〈p〉Author(s): Takumi Chikada, Hikari Fujita, Jan Engels, Anne Houben, Jumpei Mochizuki, Seira Horikoshi, Moeki Matsunaga, Masayuki Tokitani, Yoshimitsu Hishinuma, Sosuke Kondo, Kiyohiro Yabuuchi, Thomas Schwarz-Selinger, Takayuki Terai, Yasuhisa Oya〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Tritium permeation through structural materials is a critical issue in fusion reactors from the viewpoints of sufficient fuel balance and radiological hazard. Ceramic coatings have been investigated as tritium permeation barrier for several decades; however, irradiation effects of the coatings on permeation are not elucidated. In this work, yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steels by radio frequency magnetron sputtering, and their microstructures and deuterium permeation behaviors were investigated before and after iron-ion irradiation at different temperatures. An as-deposited coating had a columnar structure and transformed into a granular one after annealing. An amorphous layer formed near the coating-substrate interface of irradiated coatings, and its thickness became thinner with increasing irradiation temperature. Voids of approximately 20 nm in diameter also formed in the irradiated coatings. Deuterium permeation flux of the sample irradiated to 1 dpa at room temperature was the lowest among the unirradiated and irradiated samples, and a permeation reduction factor indicated up to 390. The amorphous layer disappeared after deuterium permeation measurements due to damage recovery, while the voids remained and aggregated. The irradiation damage would accelerate nucleation of the crystal, resulting in a decrease of the permeation flux.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2018〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 511〈/p〉 〈p〉Author(s): A. Merwin, P. Motsegood, J. Willit, M.A. Williamson〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉A parametric study of the electrolytic reduction of uranium dioxide in molten lithium chloride – lithium oxide using carbon anodes was conducted to determine the operational parameter values necessary for high yields. Operational parameters evaluated in this study include anode and cathode current densities, anodic polarization, UO〈sub〉2〈/sub〉 batch size, and method of electrochemical control. Seven oxide reduction experiments were conducted with between 25 g and 100 g UO〈sub〉2〈/sub〉. The current density on the cathode was the critical parameter, which indicates the reduction process is kinetically controlled by cathodic reactions. High cell currents were achieved without the application of high anodic potentials by using a large anode surface area. This approach facilitates efficient reduction at high throughput without production of chlorine or other corrosive gasses.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2018〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 511〈/p〉 〈p〉Author(s): R. Neu, H. Maier, M. Balden, R. Dux, S. Elgeti, H. Gietl, H. Greuner, A. Herrmann, T. Höschen, M. Li, V. Rohde, D. Ruprecht, B. Sieglin, I. Zammuto, ASDEX Upgrade Team〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Tungsten heavy alloy (97 wt% W, 2 wt% Ni, 1 wt% Fe) was investigated as an alternative for tungsten (W) as plasma facing material. It is produced commercially by several companies and compared to bulk W it is readily machinable and considerably cheaper. In order to qualify the material for use in the divertor of the mid-size tokamak ASDEX Upgrade (AUG) dedicated laboratory investigations as well as high heat flux tests in the neutral beam facility GLADIS were performed. These investigations revealed that the thermal conductivity at high temperature is close to that of W, the magnetisation is small and saturates already at low magnetic field and the hydrogen retention is similarly low as that of W. In high heat flux tests at power densities up to 20 MWm〈sup〉−2〈/sup〉 no failure was observed up to the melting temperature (〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.gif" overflow="scroll"〉〈mrow〉〈mo〉≈〈/mo〉〈msup〉〈mrow〉〈mn〉1500〈/mn〉〈/mrow〉〈mo〉∘〈/mo〉〈/msup〉〈/mrow〉〈/math〉C) of the binder phase. Even at surface temperatures of up to 2200 °C the mechanical integrity was sustained. Mechanical tests confirm the ductile behaviour of the W heavy alloy at room temperature and finite element analyses using the aforementioned data suggest a lower tendency for cracking. The increase of the long term dose-rate resulting from the activation of Ni under neutron irradiations appears to be moderate. During the 2017 campaign more than one fifth of the AUG divertor tiles consisted of W heavy alloy. Under nominal operation conditions the tiles showed no macroscopic failure and no increased Fe/Ni influx into the plasma was detected. Even though a few tiles showed strong melting at the edges due to accidental misalignment no failure due to cracking was observed.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: 1 December 2018〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 511〈/p〉 〈p〉Author(s): A. Kenich, M.R. Wenman, R.W. Grimes〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Incorporation energies and defect equilibria in monoclinic, tetragonal and cubic phases of ZrO〈sub〉2〈/sub〉 are predicted, using density functional theory calculations, for iodine dopant concentrations between 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mn〉10〈/mn〉〈/mrow〉〈mrow〉〈mo〉−〈/mo〉〈mn〉5〈/mn〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 and 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si2.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mn〉10〈/mn〉〈/mrow〉〈mrow〉〈mo〉−〈/mo〉〈mn〉3〈/mn〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 atoms per formula unit of ZrO〈sub〉2〈/sub〉. Data are presented for monoclinic and tetragonal polymorphs, in the form of Brouwer diagrams, to show the defect response at oxygen pressures ranging from 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si3.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mn〉10〈/mn〉〈/mrow〉〈mrow〉〈mo〉−〈/mo〉〈mn〉35〈/mn〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 to 10〈sup〉0〈/sup〉 atm. The oxygen pressure required for stoichiometry in monoclinic ZrO〈sub〉2〈/sub〉 is approximately 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si4.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mn〉10〈/mn〉〈/mrow〉〈mrow〉〈mo〉−〈/mo〉〈mn〉7.5〈/mn〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 atm, at both low and high iodine concentrations, whereas for tetragonal ZrO〈sub〉2〈/sub〉, it increases from 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si5.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mn〉10〈/mn〉〈/mrow〉〈mrow〉〈mo〉−〈/mo〉〈mn〉10〈/mn〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 to 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si6.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mn〉10〈/mn〉〈/mrow〉〈mrow〉〈mo〉−〈/mo〉〈mn〉6.5〈/mn〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 atm as the iodine concentration is increased from 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mn〉10〈/mn〉〈/mrow〉〈mrow〉〈mo〉−〈/mo〉〈mn〉5〈/mn〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 to 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si2.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mn〉10〈/mn〉〈/mrow〉〈mrow〉〈mo〉−〈/mo〉〈mn〉3〈/mn〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 atoms/formula unit. The dominant defects in monoclinic ZrO〈sub〉2〈/sub〉 are 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si7.gif" overflow="scroll"〉〈mrow〉〈msubsup〉〈mrow〉〈mtext〉I〈/mtext〉〈/mrow〉〈mrow〉〈mtext〉O〈/mtext〉〈/mrow〉〈mrow〉〈mo〉•〈/mo〉〈/mrow〉〈/msubsup〉〈/mrow〉〈/math〉 charge-compensated by 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si8.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈msup〉〈mtext〉I〈/mtext〉〈mo〉‴〈/mo〉〈/msup〉〈/mrow〉〈mrow〉〈mtext〉Zr〈/mtext〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉 at low oxygen pressures, and a combination of 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si8.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈msup〉〈mtext〉I〈/mtext〉〈mo〉‴〈/mo〉〈/msup〉〈/mrow〉〈mrow〉〈mtext〉Zr〈/mtext〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉, 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si9.gif" overflow="scroll"〉〈mrow〉〈msubsup〉〈mrow〉〈mtext〉I〈/mtext〉〈/mrow〉〈mrow〉〈mtext〉O〈/mtext〉〈/mrow〉〈mrow〉〈mo〉•〈/mo〉〈mo〉•〈/mo〉〈mo〉•〈/mo〉〈/mrow〉〈/msubsup〉〈/mrow〉〈/math〉 and 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si10.gif" overflow="scroll"〉〈mrow〉〈msubsup〉〈mrow〉〈mtext〉I〈/mtext〉〈/mrow〉〈mrow〉〈mtext〉i〈/mtext〉〈/mrow〉〈mrow〉〈mo〉•〈/mo〉〈/mrow〉〈/msubsup〉〈/mrow〉〈/math〉 defects at high oxygen pressures. In tetragonal ZrO〈sub〉2〈/sub〉, the dominant defects at low oxygen pressures are 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si11.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mtext〉e〈/mtext〉〈mo〉′〈/mo〉〈/msup〉〈/mrow〉〈/math〉, 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si12.gif" overflow="scroll"〉〈mrow〉〈msubsup〉〈mrow〉〈mtext〉V〈/mtext〉〈/mrow〉〈mrow〉〈mtext〉O〈/mtext〉〈/mrow〉〈mrow〉〈mo〉•〈/mo〉〈mo〉•〈/mo〉〈/mrow〉〈/msubsup〉〈/mrow〉〈/math〉 and 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si7.gif" overflow="scroll"〉〈mrow〉〈msubsup〉〈mrow〉〈mtext〉I〈/mtext〉〈/mrow〉〈mrow〉〈mtext〉O〈/mtext〉〈/mrow〉〈mrow〉〈mo〉•〈/mo〉〈/mrow〉〈/msubsup〉〈/mrow〉〈/math〉. At high oxygen pressures, 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si13.gif" overflow="scroll"〉〈mrow〉〈msup〉〈mrow〉〈mtext〉h〈/mtext〉〈/mrow〉〈mrow〉〈mo〉•〈/mo〉〈/mrow〉〈/msup〉〈/mrow〉〈/math〉 and 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si8.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈msup〉〈mtext〉I〈/mtext〉〈mo〉‴〈/mo〉〈/msup〉〈/mrow〉〈mrow〉〈mtext〉Zr〈/mtext〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉 are dominant, with additional charge-compensation from 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si14.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈msup〉〈mtext〉V〈/mtext〉〈mrow〉〈mo〉″″〈/mo〉〈/mrow〉〈/msup〉〈/mrow〉〈mrow〉〈mtext〉Zr〈/mtext〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉 defects when iodine concentrations are low. The concentration of I〈sub〉O〈/sub〉 defects in the tetragonal phase decrease with increasing oxygen pressure above stoichiometry, demonstrating competition between iodine and oxygen for occupancy of the anion site. This has implications for fuel and cladding designs that are resistant to iodine-SCC.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): Joonho Moon, Sungyu Kim, Won Dong Park, Tae Yong Kim, Samuel Westcott McAlpine, Michael P. Short, Ji Hyun Kim, Chi Bum Bahn〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Accident-tolerant fuel (ATF) cladding with high oxidation resistance during severe accidents is of critical importance to light water reactor safety and sustainability. One newly proposed ATF cladding concept, a multi-metallic layered composite (MMLC), hinges upon the oxidation resistance of an outer Fe-Cr-Si layer on top of a Zr-based alloy, separated by barrier layers to avoid Fe-Zr eutectic formation. The initial oxidation resistance of three potential Fe-Cr-Si alloys was evaluated by exposing them to 1200 °C oxidizing steam for up to one hundred seconds, along with a Zr–Nb–Sn alloy as a reference. The oxidation resistance of Fe12Cr2Si and Fe16Cr2Si was poor, exhibiting a porous, incomplete multilayer oxide composed mainly of mixed Fe/Cr/Si spinels. However, Fe20Cr2Si showed excellent oxidation resistance due to a continuous amorphous SiO〈sub〉2〈/sub〉 layer formed at the metal–oxide interface, followed by almost fully dense Cr〈sub〉2〈/sub〉O〈sub〉3〈/sub〉. This motivates the consideration of Fe-Cr-Si alloys as an additional ATF design choice, similar to FeCrAl alloys in performance and oxidation resistance mechanism.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): X.C. Meng, C. Xu, G.Z. Zuo, M. Huang, K. Tritz, D. Andruczyk, Z. Sun, W. Xu, Y.Z. Qian, J.J. Huang, X. Gao, B. Yu, J.G. Li, J.S. Hu, Huiqiu Deng〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Copper (Cu) materials are extensively used as heat sinks and sealing gaskets in fusion devices because they have the properties of good thermal and electrical conductivity and high plasticity. Meanwhile, liquid lithium (Li) is considered as a potential blanket coolant and tritium breeder and/or plasma facing material in fusion devices. Studying the corrosion characteristics of Cu materials by liquid Li under extreme fusion conditions is important because the corrosion of Cu by liquid Li may affect simultaneous application of these materials in fusion devices. The corrosion behavior of Cu in static liquid Li at 620 K for 15 h under high vacuum was investigated. After exposure to liquid Li, the weight loss rate of Cu in liquid Li is 466.1 g m〈sup〉−2〈/sup〉 h〈sup〉−1〈/sup〉, which is equivalent to 458.7 mm⋅a〈sup〉−1〈/sup〉 for the average corrosion depth rate. The entire surface of each specimen was seriously damaged. Visible grain boundary corrosion was observed on the surface of the specimens. Also, Cu debris entered the liquid Li from the corroded surface, resulting in considerable Cu loss from the specimen. These results demonstrate a corrosion protection grade of Cu in liquid Li of 10, Cu cannot withstand the corrosion of liquid Li under the given conditions. Additionally, the corrosion process of Cu in liquid Li at 620 K under high vacuum was studied using the Large-scale Atomic/Molecular Massively Parallel Simulator (LAMMPS). The results from these simulations indicate that the corrosion of Cu in liquid Li is induced via physical dissolution and intergranular corrosion, where intergranular corrosion is the dominant mechanism.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): Jianhua Ding, Pengbo Zhang, Dan Sun, Yuanyuan Wang, Shaosong Huang, Jijun Zhao〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The formation of helium-vacancy complexes under irradiation is crucial for the nucleation and early-stage growth of helium bubbles in reduced activation steels. The energetics of He〈sub〉〈em〉n〈/em〉〈/sub〉V〈sub〉〈em〉m〈/em〉〈/sub〉 (V represents vacancy) complexes (〈em〉n〈/em〉 and 〈em〉m〈/em〉 = 0–4) in Fe-9Cr alloy models and bcc Fe were investigated by first-principles calculations. A stronger self-trapping of He in Fe-9Cr alloys than in bcc Fe is found. The existence of Cr suppresses multiple He trapping in the vacancy to some extent. Besides, alloying element Cr strengthens lattice expansion induced by He atoms. Lower formation energy of He〈sub〉〈em〉n〈/em〉〈/sub〉V〈sub〉〈em〉m〈/em〉〈/sub〉 complexes and stronger binding of a He atom to He〈sub〉〈em〉n〈/em〉〈/sub〉V〈sub〉〈em〉m〈/em〉〈/sub〉 complexes in Fe-9Cr alloys as compared to in bcc Fe indicate that density of these small complexes can be higher in alloys. When 〈em〉m〈/em〉/〈em〉n〈/em〉 〉 1, aggregation of a vacancy to He〈sub〉〈em〉n〈/em〉〈/sub〉V〈sub〉〈em〉m〈/em〉〈/sub〉 complexes becomes more difficult in Fe-9Cr alloys than in bcc Fe. It is well known that the irradiation induced void swelling is related to the aggregation of small vacancy clusters or He〈sub〉〈em〉n〈/em〉〈/sub〉V〈sub〉〈em〉m〈/em〉〈/sub〉 complexes. The results indicate that the addition of Cr atoms to the Fe matrix is beneficial to the irradiation swelling resistance aided by dispersed HeV complexes.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): Yangzhong Li〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉An empirical molecular dynamics potential in the Charge-Optimized Many-Body (COMB) formalism that covers the whole uranium-oxygen composition range has been developed. Extended from a previous potential for uranium metal, this universal U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉O potential is able to model more than 20 phases of uranium oxides. The potential's flexibility, accuracy and transferability have been fully verified by rigorous testing and comparison with ab-initio calculations and experimental measurements. It is shown to be one of the most versatile and high-quality UO〈sub〉2〈/sub〉 potentials, the first potential for U〈sub〉4〈/sub〉O〈sub〉9〈/sub〉, U〈sub〉3〈/sub〉O〈sub〉7〈/sub〉 and UO〈sub〉3〈/sub〉, and the first usable U〈sub〉3〈/sub〉O〈sub〉8〈/sub〉 potential. Many important properties of major oxides in the U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉O phase diagram have been calculated and critically reviewed, including the cohesive energy, formation/reaction energies, lattice parameters, elastic constants, bulk/shear moduli, and energies for non-stoichiometric point defects and stoichiometric defects pairs. Due to its special design and parameterization process, this U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉O potential is shown to outperform all other existing ones by either obtaining higher accuracy for many of these quantities, or exclusively being able to calculate some of them. The successfully development of this potential provides a useful, reliable and convenient tool for molecular dynamics simulations that are previously impossible or unreliable to do for many materials in the U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉O system. Correction for several published oxide structures is also included in the appendix.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 30 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Beng Thye Tan, Aleksej J. Popel, Richard J. Wilbraham, Jason Day, Giulio I. Lampronti, Colin Boxall, Ian Farnan〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The escape of radionuclides from underground spent nuclear fuel disposal facilities will likely result from anoxic dissolution of spent nuclear fuel by intruding groundwater. Anoxic dissolution of various forms of uranium dioxide (UO〈sub〉2〈/sub〉), namely bulk pellet, powder and thin film, has been investigated. Long-duration static batch dissolution experiments were designed to investigate the release of uranium ions in deionized water and any surface chemistry that may occur on the UO〈sub〉2〈/sub〉 surface. The dissolved uranium concentration for anoxic dissolution of nearly stoichiometric UO〈sub〉2〈/sub〉 was found to be of the order of 10〈sup〉−9〈/sup〉 mol/l for the three different sample types. Further, clusters (∼500 nm) of homogenous uranium-containing precipitates of ∼20–100 nm grains were observed in thin film dissolution experiments. Such a low solubility of UO〈sub〉2〈/sub〉 across sample types and the observation of secondary phases in deionized water suggest that anoxic UO〈sub〉2〈/sub〉 dissolution does not only occur through a U(IV)〈sub〉(solid)〈/sub〉 to U(VI)〈sub〉(aqueous)〈/sub〉 process. Thus, we propose that dissolution of uranium under anoxic repository conditions may also proceed via U(IV)〈sub〉(solid)〈/sub〉 to U(IV)〈sub〉(aqueous)〈/sub〉, with subsequent U(IV) 〈sub〉(precipitates)〈/sub〉 in a less defective form. Quantitative analysis of surface-sensitive EBSD diffractograms was conducted to elucidate lattice-mismatch induced cracks observed in UO〈sub〉2〈/sub〉 thin film studies. Variable temperature anoxic dissolution was conducted, and no increased uranium concentration was observed in elevated temperatures.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 28 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): D.L. Porter, B.D. Miller, B.A. Hilton, M.M. Jones〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The irradiation-induced void swelling of HT9 stainless steel has been observed many times, but there are little data available to quantify the amount of swelling expected as a function of neutron exposure and temperature. This is further complicated by potential effects of heat-to-heat chemical variations or heat treatment conditions during fabrication. This paper summarizes the results of most of the relevant previously completed studies, then focuses on the swelling behavior of HT9 mixed-oxide fuel cladding and wrapper wire irradiated to high neutron exposures. Note that the hexagonal duct used for the ACO-3 experiment (in the Fast Flux Test Facility, or FFTF) was also made from HT9 and has been studied previously by other researchers. The cladding enables a wider range of operating temperatures. Immersion density and transmission electron microscopy were used to identify a peak in the amount of swelling located axially, just below core centerline, for both the cladding and the wire. The peak roughly coincided with a peak seen in the cladding diameter profile. The wrapper wire showed enhanced swelling compared to the cladding, likely due to over-tempering during wire fabrication heat treatment. The grains showed decreased martensitic structure and were larger in size. These data do not provide for the means to fully characterize the swelling of HT9, but do provide further insight into the factors controlling it. The onset of breakaway swelling was not indicated in this data.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 28 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): M. Gestin, M. Mermoux, O. Coindreau, C. Duriez, M. Pijolat, V. Peres, L. Favergeon〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Since the Fukushima Daiichi accident, increased attention is paid to the vulnerability of Spent Fuel Pools (SFPs). In case of an accidental dewatering of the fuel assemblies, the fuel cladding would be exposed to an air-steam atmosphere and its oxidation is a key phenomenon since it drives the fuel assembly heat-up and degradation. In this study, we have investigated the corrosion kinetics of pre-oxidized and as-received Zircaloy-4 (Zy-4) plate samples at 850〈sup〉∘〈/sup〉〈em〉C〈/em〉. The low temperature pre-oxidation aims at simulating the corrosion scale that grows during the in-reactor use of the fuel. High temperature oxidation tests were carried out under mixed oxygen-steam-nitrogen atmospheres. In the different atmospheres investigated, a rather protective effect of the pre-oxide scale regarding subsequent high temperature oxidation has been observed, for limited time periods however. Post-test examinations of the samples demonstrated that the loss of the protection was associated to the spalling of the pre-oxide scale that initiated at sample edges, where the pre-oxide scale was cracked. For the steam partial pressure range investigated in this study (0–8 vol%), there was no noticeable effect of the steam partial pressure on the oxidation rate. Nevertheless, samples hydrogen pick-up were strongly correlated to steam partial pressures. Moreover, 〈sup〉18〈/sup〉O isotopic labelling experiments suggested that the contribution of O〈sub〉2〈/sub〉 and H〈sub〉2〈/sub〉O to the oxidation process corresponds to their respective concentration in the gas phase.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 29 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Roman Voskoboinikov〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉A molecular dynamics study of radiation damage created in collision cascades in D0〈sub〉19〈/sub〉 Ti〈sub〉3〈/sub〉Al intermetallic compound and a Ti〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉Al binary disordered solid solution with the same chemical composition is carried out. Collision cascades are initiated by either Al or Ti primary knock-on atoms (PKA) with PKA energy 5 keV 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.gif" overflow="scroll"〉〈mrow〉〈mo〉≤〈/mo〉〈msub〉〈mrow〉〈mi〉E〈/mi〉〈/mrow〉〈mrow〉〈mi〉P〈/mi〉〈mi〉K〈/mi〉〈mi〉A〈/mi〉〈/mrow〉〈/msub〉〈mo〉≤〈/mo〉〈/mrow〉〈/math〉 20 keV in the two materials at temperature 〈em〉T〈/em〉 ranging from 100 K to 900 K. A series of at least 32 different cascades for each (〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si2.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈mi〉E〈/mi〉〈/mrow〉〈mrow〉〈mi〉P〈/mi〉〈mi〉K〈/mi〉〈mi〉A〈/mi〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉, 〈em〉T〈/em〉) set was simulated in order to imitate an isotropic spatial and random temporal distribution of PKAs, generate representative sampling and obtain statistically reliable quantitative results. The numbers of Frenkel pairs, 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si3.gif" overflow="scroll"〉〈mrow〉〈mo〉〈〈/mo〉〈msubsup〉〈mrow〉〈mi〉N〈/mi〉〈/mrow〉〈mrow〉〈mi〉F〈/mi〉〈mi〉P〈/mi〉〈/mrow〉〈mrow〉〈msub〉〈mrow〉〈mi〉α〈/mi〉〈/mrow〉〈mrow〉〈mn〉2〈/mn〉〈/mrow〉〈/msub〉〈/mrow〉〈/msubsup〉〈mo〉〉〈/mo〉〈/mrow〉〈/math〉 and 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si4.gif" overflow="scroll"〉〈mrow〉〈mo〉〈〈/mo〉〈msubsup〉〈mrow〉〈mi〉N〈/mi〉〈/mrow〉〈mrow〉〈mi〉F〈/mi〉〈mi〉P〈/mi〉〈/mrow〉〈mrow〉〈mi〉α〈/mi〉〈/mrow〉〈/msubsup〉〈mo〉〉〈/mo〉〈/mrow〉〈/math〉, formed in 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si5.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mtext〉D0〈/mtext〉〈mrow〉〈mn〉19〈/mn〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉 Ti〈sub〉3〈/sub〉Al intermetallics and Ti-25at.%Al disordered solid solution, respectively, were averaged over collision cascades with the same (〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si2.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈mi〉E〈/mi〉〈/mrow〉〈mrow〉〈mi〉P〈/mi〉〈mi〉K〈/mi〉〈mi〉A〈/mi〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉, 〈em〉T〈/em〉) and used to quantify the radiation resistance against primary damage formation. It is shown that the relationship 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si6.gif" overflow="scroll"〉〈mrow〉〈mo〉〈〈/mo〉〈msubsup〉〈mrow〉〈mi〉N〈/mi〉〈/mrow〉〈mrow〉〈mi〉F〈/mi〉〈mi〉P〈/mi〉〈/mrow〉〈mrow〉〈msub〉〈mrow〉〈mi〉α〈/mi〉〈/mrow〉〈mrow〉〈mn〉2〈/mn〉〈/mrow〉〈/msub〉〈/mrow〉〈/msubsup〉〈mo〉〉〈/mo〉〈mo〉≤〈/mo〉〈mo〉〈〈/mo〉〈msubsup〉〈mrow〉〈mi〉N〈/mi〉〈/mrow〉〈mrow〉〈mi〉F〈/mi〉〈mi〉P〈/mi〉〈/mrow〉〈mrow〉〈mi〉α〈/mi〉〈/mrow〉〈/msubsup〉〈mo〉〉〈/mo〉〈/mrow〉〈/math〉 is satisfied under all simulation conditions, 〈em〉i.e.〈/em〉 D0〈sub〉19〈/sub〉 ordered crystal structure of Ti〈sub〉3〈/sub〉Al intermetallics is an important feature affecting its radiation tolerance. Nevertheless, a relatively high resistance against the formation of primary radiation defects has been retained even after complete disordering of D0〈sub〉19〈/sub〉 Ti〈sub〉3〈/sub〉Al, which points out to the existence of additional mechanisms governing its radiation tolerance.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 24 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Jae-Hwan Kim, Masaru Nakamichi〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Beryllium intermetallic compounds (beryllides) are well-known refractory and functional materials for fission and fusion applications, specifically as reflectors and multipliers. For advanced neutron multipliers, our research group has investigated many kinds of beryllides. Due to its nuclear properties, single phase Be〈sub〉13〈/sub〉Zr beryllide was selected and successfully fabricated via plasma sintering of the homogenized Be〈sub〉13〈/sub〉Zr powder. Its reactivity against 15% H〈sub〉2〈/sub〉O was evaluated, and unexpected catastrophic oxidation (pest reaction) occurred at a low temperature range, from 973 to 1123 K, leading to the disintegration into a powder. This is caused by the intergranular degradation on the boundaries between the oxide layer and the matrix, as well as the continuous oxidation through cracks that arise from stress generation on the oxides. However, this pest reaction was suppressed by Si doping. Si likely relieves the stress near the grain boundary and in the oxide layer. The Si-doped Be〈sub〉13〈/sub〉Zr had the lowest hydrogen generation rate among the evaluated samples at 1273 and 1473 K, which is also much lower by approximately two or three orders of magnitude than that of Be at 1273 K.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 25 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Benjamin Maier, Hwasung Yeom, Greg Johnson, Tyler Dabney, Jorie Walters, Peng Xu, Javier Romero, Hemant Shah, Kumar Sridharan〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The development of a cold spray process for the deposition of chromium (Cr) coatings on zirconium-alloys is presented with the goal of improving the accident tolerance of light water reactor (LWR) fuel cladding tubes. The cold spray parameters and feedstock powders were varied to attain the desired coating properties such as thickness, microstructure, and oxidation resistance, on both Zircaloy-4 flat specimens and Optimized ZIRLO™ cladding tubes. The coated samples were tested at temperatures up to 1300 °C in air to investigate the oxidation performance and inter-diffusion between the Cr coatings and the underlying zirconium-alloy substrate. To simulate the performance of the coatings under normal LWR operating conditions, the coated samples were also tested in a steam autoclave at 400 °C and 10.3 MPa. Microstructures, phases, and hardnesses of the feedstock powders and as-deposited coatings were examined, and oxidation and inter-diffusion profiles were quantified in post-oxidation test samples. Overall, cold sprayed Cr coatings show significant promise for enhancing the accident tolerance of zirconium-alloy fuel cladding in LWRs both in terms of performance and cost-effective manufacturability.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: June 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 519〈/p〉 〈p〉Author(s): O. Pinet, J.-F. Hollebecque, I. Hugon, V. Debono, L. Campayo, C. Vallat, V. Lemaitre〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Vitrification has been selected in France as the immobilization process for high-level waste arising from spent fuel reprocessing. Several high-level waste solutions from the reprocessing of legacy UMo spent fuel, used in gas cooled reactors, have been stored in the Orano La Hague facility in stainless steel tanks since the mid-1960s. A special glass-ceramic formulation has been developed and qualified through lab and pilot testing to meet standard waste acceptance criteria for the final disposal of the UMo waste. These solutions are very rich in molybdenum and phosphorus whose contents make the molten glass quite corrosive and require a high-temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum oxide). Molybdenum is known to be sparingly soluble in conventional borosilicate glass. The formulated glass-ceramic matrix comprises a major vitreous phase containing secondary phase particles less than 100 μm in diameter. These are formed by phase separation and crystallization as the molten glass cools. The physical and microstructural properties of the UMo glass in the solid and liquid states are reported here. Evolution of microstructure as a function of the cooling profile was investigated, given the sensitivity of the crystallization process to the cooling profile. The chemical durability of the UMo glass-ceramic was also studied. The feasibility of this process has been demonstrated in a full-scale pilot facility with inactive surrogate solutions.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 30 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Romain Perriot, Christopher Matthews, Michael Cooper, Blas P. Uberuaga, Christopher R. Stanek, David A. Andersson〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Diffusion of Xe atoms in UO〈sub〉2〈/sub〉 fuel is important for nuclear fuel performance, and is enabled by interaction with U and O vacancies. Previous work using atomistic calculations based on density functional theory (DFT) and empirical potentials (EP) focused on the role of small vacancy clusters (Xe〈sub〉U2O〈em〉y〈/em〉〈/sub〉, 〈em〉y〈/em〉 = 0, 1, 2) for Xe transport, but the model predictions failed to reproduce the experimental observations by consistently underestimating Xe diffusivity. In this work, where we focused on out-of-pile conditions (i.e. in the absence of irradiation), we have explored two of the uncertainties associated with the model: the DFT methodology and the types of clusters considered. We found that using the energy barriers obtained with GGA + 〈em〉U〈/em〉 DFT allows to reconcile simulation results and experimental observations in the intrinsic regime (i.e. out-of-pile conditions). The Xe〈sub〉UO〈/sub〉 cluster is the preferred configuration at high temperature, while the mobile Xe〈sub〉U2O〈/sub〉 cluster takes over below 1720 K. The latter cluster is also the main contributor to Xe diffusion and good agreement is obtained with current models empirically fitted to experiments for diffusion at high temperature. A simple expression that captures most of Xe diffusivity is proposed, relying on Xe〈sub〉UO〈/sub〉 and Xe〈sub〉U2O〈/sub〉 clusters only. We also determined the out-of-pile concentration and diffusivity of extended clusters 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si1.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈mtext〉Xe〈/mtext〉〈/mrow〉〈mrow〉〈msub〉〈mrow〉〈mtext〉U〈/mtext〉〈/mrow〉〈mrow〉〈mi〉x〈/mi〉〈/mrow〉〈/msub〉〈msub〉〈mrow〉〈mtext〉O〈/mtext〉〈/mrow〉〈mrow〉〈mi〉y〈/mi〉〈/mrow〉〈/msub〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉, 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si2.gif" overflow="scroll"〉〈mrow〉〈mn〉3〈/mn〉〈mo〉≤〈/mo〉〈mi〉x〈/mi〉〈mo〉≤〈/mo〉〈mn〉9〈/mn〉〈/mrow〉〈/math〉, and determined that these do not contribute significantly to the overall diffusivity under intrinsic conditions. However, we speculate that the 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si3.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈mtext〉Xe〈/mtext〉〈/mrow〉〈mrow〉〈msub〉〈mrow〉〈mtext〉U〈/mtext〉〈/mrow〉〈mrow〉〈mn〉4〈/mn〉〈/mrow〉〈/msub〉〈msub〉〈mrow〉〈mtext〉O〈/mtext〉〈/mrow〉〈mrow〉〈mn〉3〈/mn〉〈/mrow〉〈/msub〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉 and 〈math xmlns:mml="http://www.w3.org/1998/Math/MathML" altimg="si4.gif" overflow="scroll"〉〈mrow〉〈msub〉〈mrow〉〈mtext〉Xe〈/mtext〉〈/mrow〉〈mrow〉〈msub〉〈mrow〉〈mtext〉U〈/mtext〉〈/mrow〉〈mrow〉〈mn〉8〈/mn〉〈/mrow〉〈/msub〉〈msub〉〈mrow〉〈mtext〉O〈/mtext〉〈/mrow〉〈mrow〉〈mn〉9〈/mn〉〈/mrow〉〈/msub〉〈/mrow〉〈/msub〉〈/mrow〉〈/math〉 clusters, which have relatively low migration barrier, may play a role under irradiation conditions, because the radiation-induced enhanced vacancy concentration would rapidly increase the fraction of large Xe-vacancy clusters.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 29 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Anamul H. Mir, Jonathan A. Hinks, Stephen E. Donnelly〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉This work explores the behaviour of xenon precipitates in amorphous silica using a transmission electron microscope with in-situ ion implantation. The specimens were first implanted at high-temperature to form equilibrium Xe precipitates which were then cooled to room temperature to form under-pressurized precipitates. In-situ implantation and real-time monitoring at high and room temperature were used to study the behaviour of the equilibrium and under-pressurized precipitates, respectively. The study at high-temperature revealed that the precipitates grow under equilibrium conditions until saturation is reached. Subsequent to precipitate growth under equilibrium conditions, the specimens contain a mixture of equilibrium, under-pressurized and possibly over-pressurized precipitates in addition to voids. Unlike precipitates growth at high-temperature (873 K), under-pressurized precipitates; formed after cooling the specimen implanted at 873 K to room temperature, considerably shrank in size when subjected to further ion implantation. The shrinkage continued until a new equilibrium state defined by the room temperature density of the Xe precipitates was achieved. We discuss the growth and shrinkage of the precipitates in terms of the ballistic thermal spike which initiates Xe diffusion from the matrix into the precipitates at high-temperature and convective flow of the glass towards the under-pressurized precipitates and voids causing their shrinkage at room temperature.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 28 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Libor Kovarik, Arun Devaraj, Curt Lavender, Vineet Joshi〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Impurity phases in metallic nuclear materials can critically influence the microstructural evolution and mechanical properties, making it crucial to understand their structure, composition, and distribution. Using transmission electron microscopy, atom probe tomography, and ab initio calculations, we provide the atomic-scale crystallographic structural and compositional analysis of an impurity phase, U〈sub〉2〈/sub〉MoSi〈sub〉2〈/sub〉C, in an important nuclear fuel U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉10Mo alloy. We identify this phase as having tetragonal symmetry with lattice parameters of 〈em〉a〈/em〉 = 6.67 Å and 〈em〉c〈/em〉 = 4.33 Å, and space group P4/mbm (No.127). Ab initio calculations were performed to verify structural stability and to perform atom position refinement.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311518317239-fx1.jpg" width="500" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: Available online 27 March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Martin Leblanc, Gilles Leturcq, Eléonore Welcomme, Xavier Deschanels, Thibaud Delahaye〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉In the framework of generation IV development for nuclear reactors, actinide mixed oxides are considered for multirecycling plutonium fuels and for transmutation targets of minor actinides. In this context, new processes are being developed for either the synthesis of mixed uranium-plutonium oxide compounds for MOx fuel or uranium-americium target fabrications. The main purposes are to simplify and step up industrial processes as well as to decrease actinide dust dispersion, and liquid effluent and gaseous releases. Among options for conversion route, a novel and patented advanced thermal denitration in presence of organic additives was established successfully to synthesize UO〈sub〉2+δ〈/sub〉, U〈sub〉0.55〈/sub〉Pu〈sub〉0.45〈/sub〉O〈sub〉2±δ,〈/sub〉 and U〈sub〉0.9〈/sub〉Am〈sub〉0.1〈/sub〉O〈sub〉2-δ〈/sub〉 oxides. Here, we describe the different intermediate steps of this process together with the characterization of the oxides obtained. The data highlight several advantages of this new route for actinide conversion.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311519302016-fx1.jpg" width="308" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: February 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 514〈/p〉 〈p〉Author(s): Tiffany C. Kaspar, Christina L. Arendt, Derek L. Neal, Shawn L. Riechers, Crystal Rutherford, Alan Schemer-Kohrn, Steven R. Spurgeon, Lucas E. Sweet, Vineet V. Joshi, Curt A. Lavender, Rick W. Shimskey〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The design of monolithic U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉Mo fuel elements fabricated from low-enriched uranium for use in high-power research reactors requires bonding of the fuel foil to either Al cladding or a Zr barrier layer. Processing of the U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉Mo ingot to final foil form has the potential to generate surface layers on the foil that differ from the bulk, metallic U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉Mo. The interfacial properties between the U〈img src="https://sdfestaticassets-eu-west-1.sciencedirectassets.com/shared-assets/16/entities/sbnd"〉Mo and Zr or Al cladding layers will then be determined by these surface layers. We use x-ray photoelectron spectroscopy, cross-sectional scanning electron microscopy, and atomic force microscopy to characterize the composition, oxidation state, and morphology of the surface layers that form after hot rolling and cold rolling depleted U–10 wt% Mo alloy (DU10Mo). A thick uranium nitride layer is observed after hot rolling, although its origin is likely from a previous processing step. The efficacy of acid etching in HNO〈sub〉3〈/sub〉 is compared to that of electropolishing in H〈sub〉2〈/sub〉SO〈sub〉4〈/sub〉 to remove surface nitride and oxide layers, and both methods are found to be similarly effective. Both laboratory (low humidity) air exposure and longer rinse times in water are shown to promote the formation of surface oxide layers. Exposure of both acid-etched and electropolished DU10Mo foils to humid air (97% relative humidity) for six weeks results in formation of a thick oxide layer due to corrosion. The oxide layer on the acid-etched foil is thicker and more highly oxidized than the oxide layer that forms on the electropolished foil, and these differences in oxidation behavior are attributed to higher surface roughness on the acid-etched foil. In general, Mo is found to play a role as a sacrificial element, typically exhibiting a larger ratio of Mo〈sup〉6+〈/sup〉/Mo〈sup〉4+〈/sup〉 than U〈sup〉6+〈/sup〉/U〈sup〉4+〈/sup〉. This is unexpected, given the greater thermodynamic driving force to form U oxides than Mo oxides.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: February 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 514〈/p〉 〈p〉Author(s): Ch. Jagadeeswara Rao, P. Venkatesh, S. Ningshen〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Ferritic chromoly steels are being considered for the process crucibles for the electrorefining of the pyrochemical reprocessing of spent metallic fuels in India. Pyrochemical reprocessing uses molten LiCl-KCl salt at 500–600 °C under an inert argon atmosphere for the electrorefining. The corrosion assessment of the 9Cr-1Mo steel in LiCl-KCl molten salt at 500 °C under inert argon atmosphere has been attempted using electrochemical measurements such as open circuit potential (OCP), linear polarization resistance and electrochemical impedance spectroscopy (EIS). The OCP of the sample was shifted towards noble direction during the total duration of 98 h. The linear polarization resistance, measured at regular intervals of exposure of 9Cr-1Mo steel to molten salt, increased with the increase of the exposure duration. The electrochemical impedance technique has been used to evaluate and monitor the molten salt-corrosion processes of the base metal. The EIS results showed the formation of a large semi-circle, with the characteristics of double loops capacitance attributed to the nature of intermittent oxide film over the surface. The surface oxide films are mostly composed of iron and chromium oxides confirmed by XRD and SEM-EDS analysis. It is also evident from the EDS analysis that depletion and enrichment of Cr and Mo across the surface of the 9Cr-1Mo steel. This work indicated that the electrochemical techniques show considerable promise for the monitoring of high-temperature molten salt corrosion processes.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): Hyung-Ha Jin, Seong Sik Hwang, Min Jae Choi, Gyeong-Geun Lee, Junhyun Kwon〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉 〈p〉We report on microstructural and mechanical property changes as a function of radiation damage value in proton-irradiated austenitic stainless steel by means of advanced characterization techniques. The microstructural changes in proton-irradiated austenitic stainless steel were analyzed by transmission electron microscopy for observation of radiation-induced defects as well as the measurement of the chemical composition at grain boundaries. The radiation hardening after the proton irradiation was characterized by nano indentation for changes in hardness profiles with radiation damage.〈/p〉 〈p〉Various transition points for microstructural and mechanical property changes under proton irradiation are analyzed via material characterization of proton-irradiated austenitic stainless steels. The saturation is expected to occur at approximately 10 displacements per atom (dpa) for the radiation-induced segregation of Cr, Ni, and P and approximately 2.5 dpa for radiation hardening. The cavity formation is observed to occur at hydrogen concentration levels greater than 5E5 atomic parts per million (appm) H. It is also found that the transition from black dot to Frank loop happened above approximately 1 dpa.〈/p〉 〈p〉Profiles of radiation-induced segregation and radiation hardening as a function of dpa can be extended to the high irradiation condition, and can be compared with experimental data for neutron irradiation-induced segregation and radiation hardening. The radiation-induced segregation after the proton irradiation at 360 °C are in good agreement with that after neutron irradiation. On the other hand, it is observed that the evolution of radiation-induced defects and the corresponding radiation hardening exhibit sooner, that appears to be because of the dose rate effect.〈/p〉 〈/div〉 〈/div〉

Description: 〈p〉Publication date: January 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 513〈/p〉 〈p〉Author(s): J. Liu, G. He, J. Hu, Z. Shen, M. Kirk, M. Li, E. Ryan, P. Baldo, S. Lozano-Perez, C. Grovenor〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉We report for the first time the observation of irradiation-induced amorphization of the zirconium suboxide formed during aqueous corrosion of Zr-0.5Nb alloys. High-resolution transmission electron microscopy results reveal amorphization of the hexagonal-ZrO suboxide under heavy ion irradiation at cryogenic temperatures. This irradiation-induced amorphization behaviour is discussed in relation to the arrangement of oxygen interstitials and the formation of stable superlattices. The sensitivity of the suboxide to irradiation damage can lead to phase changes and the accumulation of defects near the oxide/metal interface, which needs to be taken into account in the development of mechanistic models addressing radiation-assisted acceleration of corrosion rates in zirconium alloys.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311518311930-fx1.jpg" width="496" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: 15 December 2018〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 512〈/p〉 〈p〉Author(s): 〈/p〉

Description: 〈p〉Publication date: February 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 514〈/p〉 〈p〉Author(s): Zhao Shen, Junliang Liu, Koji Arioka, Sergio Lozano-Perez〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The mitigating effect introduced by intergranular Cr carbides on the stress corrosion cracking propagation of a cold-worked Alloy 600 has been firstly examined through high-resolution 3-dimensional (3D) sequential sectioning. High-resolution transmission electron microscope (TEM) and transmission Kikuchi diffraction (TKD) are used to reveal the underlying mechanisms contributing to the mitigating effect. Previously reported mechanisms contributing to the increased stress corrosion cracking resistance are evaluated and discussed. A new mechanism based on grain boundary migration inhibition and crack path deviation is proposed.〈/p〉〈/div〉 〈/div〉 〈h5〉Graphical abstract〈/h5〉 〈div〉〈p〉〈figure〉〈img src="https://ars.els-cdn.com/content/image/1-s2.0-S0022311518309735-fx1.jpg" width="418" alt="Image 1" title="Image 1"〉〈/figure〉〈/p〉〈/div〉

Description: 〈p〉Publication date: February 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 514〈/p〉 〈p〉Author(s): Zhao Shen, Kai Chen, Xianglong Guo, Lefu Zhang〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉General corrosion and stress corrosion cracking (SCC) susceptibility of an oxide dispersion strengthened (ODS) austenitic 310 (310-ODS) steel in supercritical water (SCW) are studied by weight gain and slow strain rate tensile (SSRT) testing, respectively. 310-ODS steel shows an excellent general corrosion resistance in SCW at 600 °C. Energy dispersive X-ray (EDX) and electron backscattered diffraction (EBSD) are conducted on the cross-section of the surface oxide film, revealing a double-layered structure. Results from SSRT tests at 600 °C show an intergranular fracture mode, and SCC susceptibility of 310-ODS steel increases with the increasing of dissolved oxygen (DO) concentration. SSRT tests at 650 °C show ductile fracture mode, and SCC susceptibility is minimum. ODS enhances the yield strength of 310-ODS steel to over 480 MPa, tensile strength to over 750 MPa, and still keeps elongation rate to over 12% at 600 °C. Combining with its low general corrosion rate and low SCC susceptibility, 310-ODS steel is supposed to be a promising material for the fuel cladding of supercritical water-cooled reactor (SCWR).〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: February 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 514〈/p〉 〈p〉Author(s): R.W. Harrison, G. Greaves, J.A. Hinks, S.E. Donnelly〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Intermetallic Re precipitation at concentrations below the solubility limit is a puzzling phenomenon in neutron irradiated W. Ion irradiation has been unable to reproduce this, denying the community the ability to accurately simulate neutron damage microstructures and probe precipitate formation. We have recently been successful in inducing σ (WRe) and χ (WRe〈sub〉3〈/sub〉) phase formations in W26Re irradiated with 350 keV Ne ions at 500 and 800 °C. The precipitation of these phases is related to the effects of cascade energy density and ballistic mixing during previous high energy self-ion irradiations and is concluded to have caused redissolution of precipitates and prevented their observation.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: February 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 514〈/p〉 〈p〉Author(s): Chengxu Wang, Wei Chen, Minghui Chen, Demin Chen, Ke Yang〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉CrN barrier was deposited between a Ni-electroplating and GH3535 alloy with the goal of prohibiting elements interdiffusion and increasing corrosion resistance in molten FLiNaK (LiF-NaF-KF: 46.5-11.5-42.0 mol%) salt. Results indicate that the Ni-coating apparently improves corrosion resistance of GH3535 in molten fluoride salts, but Cr and Fe diffuse constantly from the alloy to the Ni-coating. They are finally dissolved into the molten salt, causing corrosion of the GH3535 alloy. With a CrN diffusion barrier, however, elements interdiffusion has been completely suppressed and the corrosion resistance is further improved by more than eight times comparing to the CrN-free Ni-coated alloy.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: February 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 514〈/p〉 〈p〉Author(s): Vianney Motte, Dominique Gosset, Gaëlle Gutierrez, Sylvie Doriot, Nathalie Moncoffre〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉When B〈sub〉4〈/sub〉C boron carbide is irradiated in neutron reactors, high quantities of helium are produced due to the 〈sup〉10〈/sup〉B(n,α)〈sup〉7〈/sup〉Li neutron absorption reactions. It is well known that most helium remains in the material, this inducing swelling and cracking. However very few studies deal with the nucleation and growth of the helium clusters. In this paper, B〈sub〉4〈/sub〉C is implanted with helium and its behaviour is studied mainly thanks to TEM observations. Different parameters are considered: helium concentration, implantation temperature, post-implantation annealing temperature, width of the implanted zone and overlap with ballistic damage simulating the fast neutron scattering. We observe that helium clusters form either on structural defects (grain and twin boundaries, precipitates) or thanks to irradiation damage induced nucleation. When implanted at low temperature and then annealed, nanometric bubbles form that eventually agglomerate as flat discs associated with strong strain fields. In the temperature range encountered in fast neutron reactors, those discs evolve toward parallel 20–50 nm diameter helium platelets. Grain boundaries are very efficient trapping sites, no evolution of the intergranular helium bubbles is observed up to 1200 °C. In the implanted zone, the platelets change into polyhedron bubbles over 1100 °C with a relaxation of the strain fields together with a strong density decrease. The addition of a ballistic damage increases the density of the helium bubble germs.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 515〈/p〉 〈p〉Author(s): Mi Wang, Miao Song, Calvin R. Lear, Gary S. Was〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉Thirteen alloys including high- and low-strength nickel-base alloys, austenitic stainless steels, and ferritic alloys were irradiated using 2 MeV protons to a damage level of 2.5 dpa at 360 °C and assessed for their susceptibility to irradiation assisted stress corrosion cracking (IASCC) in both BWR normal water chemistry (NWC) and PWR primary water. Cracking susceptibility was highest for high strength nickel-base alloys, followed by the low strength nickel-base alloys and then the low strength iron-base alloys. Cracking in the nickel-based alloys was worst in normal water chemistry, which was reversed for the iron-based alloys. In general, cracking correlated with the degree of microstructure changes, though no single feature could be linked to cracking. IGSCC occurred in both the unirradiated and irradiated conditions in high strength nickel-base alloys with susceptibility being considerably higher following irradiation. In all cases, slip was planar, and the degree of slip localization correlated with the probability of IG crack initiation. Low strength nickel-base alloys showed the same dependence on environment as high strength alloys but were considerably less susceptible to IASCC initiation. Among the low strength iron-base alloys, alloy 800 was most susceptible to IASCC initiation in both BWR NWC and PWR primary water, which also correlated with grain boundary chromium depletion and silicon segregation. Across all alloys, cracking correlated with both the degree of localized deformation and the hardness in the irradiated condition. The agreement is expected as increased hardening also correlates with localized deformation, which is likely a necessary, though insufficient condition for cracking.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: March 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 515〈/p〉 〈p〉Author(s): Jared O. Kroll, Jarrod V. Crum, Brian J. Riley, James J. Neeway, R. Matthew Asmussen, Martin Liezers〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉This paper details synthesis methods that were used to fabricate phase-pure Nd〈sub〉3〈/sub〉BSi〈sub〉2〈/sub〉O〈sub〉10〈/sub〉 using a LiCl-flux method. In this study, a variety of conditions were evaluated to find the optimal processing parameters for maximizing the Nd〈sub〉3〈/sub〉BSi〈sub〉2〈/sub〉O〈sub〉10〈/sub〉 yield including the mass ratio of LiCl to target mass of Nd〈sub〉3〈/sub〉BSi〈sub〉2〈/sub〉O〈sub〉10〈/sub〉 (i.e., 1:3, 1:2, 1:1, 2:1, and 6:1) as well as the soak temperature (i.e., 700, 800, or 900 °C) at the 1:1 ratio. It was found that the optimal ratio of LiCl:Nd〈sub〉3〈/sub〉BSi〈sub〉2〈/sub〉O〈sub〉10〈/sub〉 was 1:1 and the optimal heat-treatment temperature was 900 °C (of those temperatures evaluated). Product made from the 1:1 material synthesized at 900 °C was recovered and fired at temperatures of 925–1350 °C. After 4-h firings at temperatures of 925–1100 °C, the Nd〈sub〉3〈/sub〉BSi〈sub〉2〈/sub〉O〈sub〉10〈/sub〉 phase purity was found to be 100% with powder X-ray diffraction, but started to decompose at the longer firing times at 1100 °C and higher temperatures (1350 °C) at the 4-h dwell time. Thus, the lowest firing temperature for achieving a phase-pure pellet was 925 °C under the conditions studied here. The phase-pure sample was found to contain 28 vol% open porosity determined by Archimedes’ method. The durability of the Nd〈sub〉3〈/sub〉BSi〈sub〉2〈/sub〉O〈sub〉10〈/sub〉 product was examined in flow-through conditions at 90 °C with contacting solution between pH〈sub〉90ºC〈/sub〉 4.1 and 10.1. The release rate of Nd from the material had a maximum at pH〈sub〉90ºC〈/sub〉 4.1 calculated to be 4.93 g m〈sup〉−2〈/sup〉 d〈sup〉−1〈/sup〉 and a minimum release rate of Nd at pH〈sub〉90ºC〈/sub〉 7.7 of 〈1.34 × 10〈sup〉−5〈/sup〉 g m〈sup〉−2〈/sup〉 d〈sup〉−1〈/sup〉. These release rates are roughly three orders of magnitude lower than what is normally measured from typical glass waste forms at pHs of 7–9.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: February 2019〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials, Volume 514〈/p〉 〈p〉Author(s): Mao Yang, Yichao Gong, Guangming Ran, Hailiang Wang, Ruichong Chen, Zhangyi Huang, Qiwu Shi, Xiaojun Chen, Tiecheng Lu, Chengjian Xiao〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉The tritium release behavior of the Li〈sub〉4〈/sub〉SiO〈sub〉4〈/sub〉 pebbles and Li〈sub〉4〈/sub〉SiO〈sub〉4〈/sub〉 + 5 mol% TiO〈sub〉2〈/sub〉 pebbles with small grain sizes was investigated. The tritium release results of Li〈sub〉4〈/sub〉SiO〈sub〉4〈/sub〉 pebbles with different grain sizes (0.3 μm and 1.5 μm) indicated that the grain size had little effect on the tritium release regardless of the difference in gas composition of purge gas. Moreover, the tritium release of small grained pebbles was dominated by the desorption process, since the addition of H〈sub〉2〈/sub〉 to purge gas substantially affected the release behavior. The modified Li〈sub〉4〈/sub〉SiO〈sub〉4〈/sub〉 pebbles with addition of TiO〈sub〉2〈/sub〉 exhibited enhanced water formation capacity due to the increased concentration of active point as the oxygen supplier. Obvious tritiated water release peaks around 690 °C could be observed from the Li〈sub〉4〈/sub〉SiO〈sub〉4〈/sub〉 + 5 mol% TiO〈sub〉2〈/sub〉 pebbles under 0.1%H〈sub〉2〈/sub〉+He purge gas. The modified Li〈sub〉4〈/sub〉SiO〈sub〉4〈/sub〉 pebbles also showed improved tritium release behavior, the tritium release peaks shifted to lower temperatures compared to Li〈sub〉4〈/sub〉SiO〈sub〉4〈/sub〉 pebbles.〈/p〉〈/div〉 〈/div〉