Improved crystallinity of GaP-based dilute nitride alloys by proton/electron irradiation and rapid thermal annealing

This study presents the positive effects of proton/electron irradiation on the crystallinity of GaP-based dilute nitride alloys. It is found that proton/electron irradiation followed by rapid thermal annealing enhances the PL peak intensity of GaPN alloys, whereas major photovoltaic III-V materials such as GaAs and InGaP degrade their crystal quality by irradiation damage. Atomic force microscopy and transmission electron microscopy reveal no degradation of structural defects. GaAsPN solar cell test devices are then fabricated. Results show that the conversion efficiency increases by proton/electron irradiation, which is mainly caused by an increase in the short-circuit current.

RADIATION INDUCED SOFTENING OF CRYSTALS

Under irradiation of crystals, atomic vibrations of the lattice that are large enough in amplitude so that the linear approximation and therefore the conventional phonon description of the lattice is not enough. At the same time, these vibrations are localized and can travel long distances in a crystal lattice [1, 2]. In metals and other crystals, they are called discrete breathers (DBs), which are the secondary products of irradiation damage, the primary one being the creations of defects that involve atom displacements to produce point and extended defects, which results in radiation induced hardening (RIH). A part of the remaining energy transforms in DBs before decaying into pho-nons. Thus, while a material is being irradiated in operational conditions, as in a reactor, a considerable amount of DBs with energies of the order of one eV is produced, which helps dislocations to unpin from pinning centers, pro-ducing Radiation Induced Softening (RIS), which opposes RIH [3, 4]. This effect is investigated under (in-situ) im-pulse and steady-state electron irradiation.

Progress on neutronic analysis for CFETR

The neutron induced irradiation field is a key problem in fusion reactor related to nuclear responses, shielding design, nuclear safety, and thermo-hydraulic analysis. To support the system design of China Fusion Engineering Test Reactor (CFETR), the comprehensive analysis of irradiation field has been conducted in support of many new developed advanced tools. The paper first summarizes the recent progress on related neutronics code development effort including the geometry conversion tool cosVMPT, Monte Carlo variance reduction technology ‘on-the-fly’ global variance reduction (GVR). Such developed tools have been fully validated and applied on the CFETR nuclear analysis. The neutron irradiation has been evaluated on CFETR Water Cooled Ceramic Breeder (WCCB) blanket, divertor, vacuum vessel, superconductive coils and four kinds of heating systems including the Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating (ICRH), Low Hybrid Wave (LHW) and Neutral Beam Injection (NBI). The nuclear responses of tritium breeding ratio (TBR), heating, irradiation damage, Hydrogen/Helium (H/He) production rate of material have been analyzed. In case of neutron damage and overheating deposition on the superconductive coils and Vacuum Vessel (VV), the interface and shielding design among heating systems, blanket and other systems has been initialized. The results show the shielding design can meet the requirement of coil and VV after several iterated neutronics calculation.

Revealing nano-scale lattice distortions in implanted material with 3D Bragg ptychography

Small ion-irradiation-induced defects can dramatically alter material properties and speed up degradation. Unfortunately, most of the defects irradiation creates are below the visibility limit of state-of-the-art microscopy. As such, our understanding of their impact is largely based on simulations with major unknowns. Here we present an x-ray crystalline microscopy approach, able to image with high sensitivity, nano-scale 3D resolution and extended field of view, the lattice strains and tilts in crystalline materials. Using this enhanced Bragg ptychography tool, we study the damage helium-ion-irradiation produces in tungsten, revealing a series of crystalline details in the 3D sample. Our results lead to the conclusions that few-atom-large ‘invisible’ defects are likely isotropic in orientation and homogeneously distributed. A partially defect-denuded region is observed close to a grain boundary. These findings open up exciting perspectives for the modelling of irradiation damage and the detailed analysis of crystalline properties in complex materials.

Actinides induced irradiation damage and swelling effect in irradiated Zircaloy-4 after 30 years of storage

After several years in the reactor core, irradiated nuclear fuel is handled and subsequently stored for a few years under water next to the core, to achieve thermal cooling and decay of very short-lived radionuclides. Thereafter, it might be sent to dry-cask interim storage before final disposal in a deep geological repository. Here, the spent nuclear fuel (SNF) is subject to a series of physicochemical phenomena which are of concern for the integrity of the nuclear fuel cladding. After moving the SNF from wet to dry storage, the temperature increases, then slowly decreases, leading the hydrogen in solid solution in the cladding to precipitate radially with consequent hydride growth and cladding embrittlement (Kim, 2020). Another phenomenon affecting the physical properties of the cladding during interim dry storage is the irradiation damage produced in the inner surface of the cladding by the alpha decay of the actinides present at the periphery of the pellet, particularly when the burnup at discharge is high. SNF pellets with high average burnup present larger fuel volumes at the end of their useful life due to accumulation of insoluble solid fission products and noble gases, which leads to disappearance of the as-fabricated pellet–clad gap. Further swelling is expected as a consequence of actinide decay and the accumulation of helium. This leads to larger cladding hoop stress and larger alpha decay damage. The present work first investigates the variation in diameter caused by pellet swelling in an irradiated Zircaloy-4 cladding after chemical digestion of the uranium oxide (UOx) pellet. Second, the irradiation damage produced during the 30 years elapsed since the end of irradiation in terms of displacements per atom (dpa) is studied by means of the FLUKA Monte Carlo code. The irradiation damage produced by the decay of actinides in the inner surface of the cladding extends for less than 3 % in depth. The considered cladded UOx pellet was extracted from a pressurized water reactor (PWR) fuel rod consisting of five segments, with an average burnup at discharge of 50.4 GWd (tHM)−1.

Physical Testing and Surface Morphology Studies of Stainless Steel Irradiated with Xe Ions

6 MeV Xe ions were used to irradiate austenitic stainless steel at room temperature. Three displacement damage levels of 2,7 and 15 dpa were selected. Microstructure and surface morphology were characterized by transmission electron microscopy (TEM), positron annihilation lifetime spectroscopy (PLS) and atomic force microscope (AFM). PLS indicated that vacancy defects were introduced by ions irradiation. Vacancy clusters containing Xe will reduce the positron annihilation lifetime. High density of dislocation loops were observed by TEM. The dislocation loops size and density saturates after 7 dpa and the nature of dislocation loops can be deduced by its distribution. A surface step was detected by AFM measurements between irradiated region (uncovered) and unirradiated region (covered with nickel mesh). This indicate that the irradiation swelling phenomenon occur and swelling is closely related to irradiation damage. According to the step height, the volume swelling is about 1.7% and 4.2% after irradiated to 7 and 15 dpa.

Microstructure characterization of reactor pressure vessel steel A508-3 irradiated by heavy ion

As one of the key structures used in nuclear power plants, the study of irradiation effects of pressure vessel steel (RPV) is of great scientific value to nuclear safety. The RPV steel was irradiated by Fe ions up to three different irradiation damage levels (0.08 dpa, 0.15 dpa, and 0.6 dpa). The transmission electron microscope was utilized to measure the irradiated microstructure and it was found that after the irradiation of 0.08 dpa, the density and size of dislocation loops in Fe ions irradiated samples was small and the dislocation loops were distributed near the surface. When irradiation dose was up to 0.15 dpa, many black dots were distributed in the whole irradiation region and some large size dislocation loops appeared. In the case of 0.6 dpa, a large number of dislocation loops were produced and the distribution of dislocation loops extended to the whole irradiation region owing to the production and growth of defects such as vacancies and black dots.