查看更多>>摘要：One of the most critical issues affecting the integrity of nuclear fuel cladding tubes is hoop stress-induced hydride reorientation from the circumferential to the radial direction. The theoretical prediction of the threshold stress for the hydride reorientation is a long-standing but unrealized objective of material design engineers in the nuclear industry. In this paper, we propose a criterion to determine the threshold stress based on the thermodynamic model. The results obtained this way agree very well with experimental observations. Multiple factors, such as crystallographic texture, grain-boundary structure and distribution, grain morphology and size, the hydride-matrix misfit strain, hydrogen content, the chemical composition, the temperature and the mechanical strength and anisotropy of the zirconium matrix, simultaneously control the difficulty level of hydride reorientation. Strengthening only the radial basal pole texture of the zirconium alloy tube is not enough to effectively enhance the threshold stress of hydride reorientation; other factors, especially the grain-boundary structure and distribution, also need to be considered. The model presented in this study elucidates the mechanism of stress reorientation of hydrides and offers new insight on how to further increase the resistance to hoop stress-induced hydride reorientation in nuclear fuel cladding tubes (c) 2022 Elsevier B.V. All rights reserved.
Ard, Johnathon C.Yingling, Jacob A.Johnson, Kaitlin E.Schorne-Pinto, Juliano...
12页
查看更多>>摘要：The Molten Salt Thermal Properties Database - Thermochemical (MSTDB -TC) has been developed to support thermodynamic modeling of fluoride-and chloride-based systems for molten salt reactors (MSRs). Utilizing data from available literature and original work, the current MSTDB -TC Ver. 1.2 contains models for 96 pseudo-binary systems, 37 pseudo-ternary systems, 42 solid solutions, 229 stoichiometric compounds, and 130 gaseous species. This includes 67 reassessed systems and 13 original system models. The reassessment of the LiCl-RbCl pseudo-binary system and the original assessment of the UF3-UF4 pseudo-binary system are included as examples of those effort s and provide improved representations of those systems. The database continues to be expanded, adding systems of relevance to the wide variety of MSR concepts. To illustrate applications for MSTDB -TC, phase equilibria were computed to understand the effect of varying uranium content in the KCl-MgCl2-NaCl -UCl3 system. In an example of how such calculations can support corrosion modeling, the control of redox potential by the UF3:UF4 ratio in the LiF-BeF2-UF3 -UF4 system is demonstrated and used to assess corrosion potential for typical alloy constituents.(C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Oxyapatite (general formula: A(2)(I)A(8)(II) (BO4)(6)O-2) is a potential material for immobilization of nuclear waste. Here, a novel Sr2Nd8(SiO4)6O(2) glass-ceramics (GCs) derived from uranium tailings were synthesized by consecutive microwave sintering technology at 1200 degrees C for simultaneous immobilization of simulated bivalent fission product (FP) Sr and trivalent actinide (An(3+)) Nd. Sr and Nd were successfully immobilized into A(2)(I) and A(8)(II) acceptor sites of oxyapatite respectively, and glass structures. The phase evolution, microstructure, physical property, and radon exhalation of as-prepared sintered forms were systematically investigated by XRD, FT-IR, SEM-EDS, TEM, and RAD7 radon meter. The experimental results demonstrated that the ultimate solid solubility of Sr reached 25 wt.%, while Nd exceeded 30 wt.% at 1200 degrees C. Importantly, the solidified samples exhibited homogenous and dense microstructure, and the radon exhalation rate improved with the increase of pore. It was indicated that oxyapatite glass-ceramics could be a potential matrix for the simultaneous immobilization of bivalent FP and An(3+). (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：This study used electron microscopy techniques and phase field modeling to investigate the mechanisms responsible for developing the high burnup structure in U-Mo fuels. The results show that grain subdivision is initiated by polygonization; however, evidence of dynamic recrystallization is present with increasing fission densities. At 2.5 x 10( 21) fissions/cm(3), LAGB misorientation of 4 & nbsp; was measured near the grain boundary of neighboring grains, which supports polygonization as the leading grain subdivision mechanism. However, with increasing fission densities, the formation of HAGB subgrain clusters are evident, marking the activation of another mechanism dynamic recrystallization. Previously formed LAGBs transform into new HAGB grains, which is reflected in the increasing number frequency of HAGBs to LAGBs. At the same time, the HAGB subgrain clusters also undergo polygonization, suggesting that polygonization is a continuous mechanism. Based on the simulations performed in this study, polygonization may continue to occur until a mean grain diameter of-382 nm is achieved. Based on the results of this work, LAGB to HAGB transformation is associated with a strain-driven mechanism potentially caused by fission gas-induced plastic deformation and is consistent with recrystallization. Polygonization and dynamic recrystallization occurs in tandem with polygonization leading grain subdivision. Based on the fission densities evaluated in this study, authors postulate that the critical fission density at which dynamic recrystallization (DRX) becomes active is between 2.5 x 10(21) fissions/cm(3) and 3.5 x 10(21)& nbsp;fissions/cm(3). (C)& nbsp;2022 Elsevier B.V. All rights reserved.
Gussev, M. N.Zhong, W.Garner, F. A.Freyer, P. D....
15页
查看更多>>摘要：AISI 304 L austenitic stainless steel is one of the major structural materials used in light water reactors (LWRs). In the future, 304 L components may require repair, and several welding techniques have been proposed as candidates. In this study, laser beam welding using the low-energy contribution approach was performed in-2016 on neutron-irradiated AISI 304 L steel with nominal values of 3 and 8 appm helium (He). The goal was to investigate the impact of helium on the irradiated material weldability and evaluate helium-associated damage. The weld and heat-affected zone (HAZ) microstructure was studied using scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy (TEM). Analysis of the weldment cross-sections did not reveal severe cracking. Still, evidence of incipient helium-associated damage was observed in the HAZ, manifesting as degraded grain boundaries (GBs) "decorated" by pore chains and mostly small (i.e., < 25-30 mu m) scattered cracks. The largest observed crack was-50 mu m in length. Helium-associated damage was localized within-20 0-30 0 mu m of the weld pool boundary, and the fraction of the compromised GBs, as a rule, was below-9-11% of the total GB network. TEM analysis showed significant annealing of the radiation-induced defects at distances up to-300 mu m from the weld pool boundary. Inside the HAZ, the cavities (likely, helium bubbles) tended to form at GBs and inclusions, such as manganese sulfide precipitates in the grain interior. Crystallography analysis of the HAZ damage showed that random high-angle boundaries were most susceptible to helium-associated damage. In contrast, low-angle random boundaries and twin boundaries appeared to be strongly resistant to degradation from the combined effects of helium and welding.(c) 2022 Elsevier B.V. All rights reserved.
Cheng, BinDuchnowski, Edward M.Sprouster, David J.Snead, Lance L....
15页
查看更多>>摘要：Engineered composites composed of a radiation stable continuous matrix containing a highly moderating entrained phase are attractive candidates for the realization of bulk materials that are structurally and neutronically superior to nuclear graphite. Here, we explore neutronics driven selection of entrained moderating phases in MgO-based ceramic composites with a focus on the MgO-BeO system given its exceptional moderating power and high temperature stability. Using lithium-bearing salts as sintering aids, fully dense MgO-BeO composites with BeO loading up to 40 vol.% are produced through direct current sintering at markedly reduced temperatures relative to phase-pure MgO. Thermophysical properties mapped as a function of the BeO concentration are shown to align with various composite models, thus revealing the influence of underlying defects on the thermophysical property trends. From microreactor neutronics and thermal hydraulic calculations, the MgO-40BeO moderator is shown to increase both cycle length and fuel utilization relative to graphite and with steady-state temperature distributions remaining within specification. The ceramic composite moderators outperform graphite for all metrics considered with significant potential demonstrated for reducing energy costs while enabling novel microreactor designs through the replacement of graphite.(c) 2022 The Author(s). Published by Elsevier B.V.This is an open access article under the CC BY-NC-ND license( http://creativecommons.org/licenses/by-nc-nd/4.0/ )
查看更多>>摘要：First-principles calculations were performed to examine the defect-induced energy storage in graphite. The accumulation of energy resulting from inducing defects in graphite is a well-known phenomenon. Given the recent interest in exploiting this process for energy-storing purposes, more careful investigation is necessary. Some of the earliest studies of damaged graphite, and the stored energy associated with that, were motivated by the associated technological issues in nuclear reactor operation. A large number of excited state defects, for example Frenkel pairs, can be generated in graphite through bombardment of high-energy neutrons. The sudden release of this energy (also called Wigner energy) poses a serious concern to the safe operation of nuclear reactors. At the same time, controlled defect generation in graphite using neutron/ion irradiation might represent a potential energy storage mechanism. In recently published papers, the design of an integrated energy system that couples a nuclear reactor with a Wigner effect-based energy storing system was proposed. To accurately estimate the performance that can be achieved in terms of stored energy density through defect generation, density functional theory (DFT) based first-principles calculations were performed. In this work, stored energy accumulation was modeled in two ways by Frenkel pair accumulation and overlapping collision cascade methods. The former was done with ab initio molecular dynamics (AIMD) simulations, and the latter was done with combined classical molecular dynamics (MD) and AIMD simulations. The agreement between the calculated and experimental results for how stored energy changes with dosage suggests that this model could be useful for the on-going research into damaged graphite as an energy storage medium.(c) 2022 Elsevier B.V. All rights reserved.
Taller, StephenPauly, ValentinHanbury, RigelWas, Gary S....
14页
查看更多>>摘要：Dual-ion irradiations using 5.0 MeV defocused Fe2+ ions and co-injected energy degraded 2.00 to 2.85 MeV He2+ ions were conducted on a Fe9CrMo ferritic-martensitic steel T91 to 17 dpa at a damage rate range of 5 x 10(-5) dpa/s to 3 x 10(-3) dpa/s at 445 degrees C, followed by characterization of the microstructure using conventional and scanning transmission electron microscopy. Radiation induced Ni/Si clusters and radiation induced segregation were quantified using energy dispersive X-ray spectroscopy at each condition and were compared with the same material irradiated in the BOR-60 reactor and in the as-received condition. No significant Cr segregation was found at lath boundaries after dual-ion irradiation, while Ni and Si enrichments both decreased with increasing damage rate leading to a sharp decrease in the density of Ni/Si clusters with damage rate. Increased point defect recombination at higher ion damage rates likely reduced the Ni/Si cluster density compared with BOR-60. Although the overall vacancy concentration and diffusion are enhanced by the irradiation damage rate, the lack of time for thermal diffusion and ballistic displacements of solutes are significant limiting factors for Ni/Si cluster formation. This work demonstrates the effect of irradiation damage rate on elemental segregation and clustering when using ion irradiation to simulate reactor irradiation. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：The UMo/Al dispersion fuel in plate form is considered for the conversion of high-performance research reactors in Europe. In the framework of the UMo fuel qualification program, the adequate margin of safety performance by considering several technological solutions associated with the fabrication parameters, such as heat-treatment of the UMo particles, coating with a diffusion barrier material and powder size distribution as some examples. All these parameters, along with the effect of the hot-rolling process were evaluated by means of image processing and detailed microstructural characterizations for fresh samples i.e. prior to irradiation tests. Principle macroscopic features of powder batches include the size and shape distributions and coating surface examinations. Microscale investigations explored both the coating and kernel microstructures as well as the interface layer between them. Finally, nanoscale analyses examined the UMo-coating interface. The extensive stresses associated with the hot-rolling process have a significant impact on the deformation of the UMo kernels and the degradation of the coated film. The UMo kernels mostly lost their spherical shape for faceted and elongated shapes whereas three types of film degradations were identified including cracks, chippings and delamination.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Alumina-forming austenitic (AFA) stainless steel with excellent properties has been regarded as a promising candidate of fuel claddings in Supercritical carbon dioxide (S-CO2) gas cooled nuclear reactor. In current work, in-situ irradiation with 400 keV Fe+ in transmission electron microscope were performed on AFA stainless steel to investigate the influences of irradiation dose and pre-existing dislocation lines on loop evolution at 773 K. The irradiation damage caused by irradiation-induced loops was also analyzed. In-situ TEM observation displayed the evolution and characteristics of dislocation loops, including initiation, migration, aggregation, growth, annihilation, and combination. Both Frank loops with b = 1/3 < 111 > and perfect loops with b = 1/2 < 110 > were formed, and at 0.54 dpa, 1/2 < 110 > loops accounted for the majority and was up to 64.6%. Pre-existing dislocation lines obviously affected not only the evolution and distribution of dislocation loops, but also the degree of irradiation hardening. At the same irradiation dose, not only the size and number density of dislocation loops in the region with dislocation lines were smaller than those in the region without dislocation lines, but also the increment of yield strength and irradiation damage was lower before the formation of dislocation network. Increasing pre-existing dislocation density should be beneficial to improve the irradiation resistance of materials, but it needed to be comprehensively considered with the mechanical properties of the material. Al component played an important role in irradiation behavior of AFA steels and the corresponding mechanism was discussed subsequently. (C) 2022 Elsevier B.V. All rights reserved.
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