Journal of Nuclear Materials最新文献

{"title":"Oxidation behavior of β-Nb formed in Zr-1Nb under neutron irradiation in PWR conditions","authors":"Xue Han,&nbsp;Huacai Wang,&nbsp;Huanlin Cheng,&nbsp;Jinze Sun,&nbsp;Lina Guo,&nbsp;Wulin Song,&nbsp;Huize Fan","doi":"10.1016/j.jnucmat.2024.155478","DOIUrl":"10.1016/j.jnucmat.2024.155478","url":null,"abstract":"<div><div>This work focuses on neutron irradiated Zr-1Nb alloy, using High Resolution Transmission Electron Microscopy (HRTEM) to investigate the oxidation behavior of β-Nb at different distances from the Oxide /Metal (O/M) interface within the oxide film. Results show that β-Nb was initially oxidized to T-NbO₂ at 0 nm at O/M interface, then into a complex morphology of T-NbO₂, M-Nb₂O₅, and O-Nb₂O₅ within 600 nm. Finally, it was completely oxidized to M-Nb₂O₅ within 800 nm. β-Nb in this study did not exhibit amorphous morphology within observed distances. In addition, Inverse Fast Fourier Transformation (IFFT) and Weak Beam Dark Field (WBDF) techniques are employed to characterize the dislocation density and distribution in the oxide film, results indicate that the distribution of dislocations generated by neutron irradiation in the oxide film is relatively uniform and neutron irradiation is not the primary reason affecting the oxidation behavior of β-Nb.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155478"},"PeriodicalIF":2.8,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652826","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Determining reference standard strength for neutron-irradiated reduced activation ferritic/martensitic steel F82H by Bayesian method","authors":"Takashi Nozawa ,&nbsp;Hiroyasu Tanigawa ,&nbsp;Taichiro Kato ,&nbsp;Xiang (Frank) Chen ,&nbsp;Yutai Katoh","doi":"10.1016/j.jnucmat.2024.155486","DOIUrl":"10.1016/j.jnucmat.2024.155486","url":null,"abstract":"<div><div>The deterministic approach widely adopted in the design of structural components relies on systematically defined design limits using empirically determined safety factors. However, this approach is not always appropriate because structures are subjected to a variety of loads in the practical environment, which may result in excessively conservative design limits. In recent years, a more rigorous probabilistic approach that incorporates material strength distributions has become an important solution. In the probabilistic approach, the probability density functions of material strength properties underpin the design criteria. The objective of this study is to identify the density distribution functions that best describe tensile properties of irradiated F82H to define a reference strength for DEMO design. Due to the limited number of existing data, this study specifically employs a Bayesian prediction method based on Monte Carlo simulations to determine a material reference value with statistical reliability and to investigate its effectiveness. For example, the dependence of tensile properties of 300 °C irradiated materials on irradiation damage and the range predicted by 95% Bayesian estimation was evaluated. As a statistical model for the dose dependence of statistical parameters, the normal distribution exhibited a better fit for 0.2% proof strength and tensile strength, whereas the distribution of total elongation data gave comparable reference values for both the normal and Weibull distribution models. Both models gave comparable criteria for the distribution of total elongation data. The Weibull model also gave better results for uniform elongation. The function best describing the model was a logarithmic law for both 0.2% proof strength and tensile strength, while a power law for both total and uniform elongation, which allowed for more comprehensive data prediction of irradiation data with statistical accuracy for DEMO reactor design.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155486"},"PeriodicalIF":2.8,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571578","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Role of severe plastic deformation on mechanical behavior of irradiated materials: A case study with Nb-1Zr alloy","authors":"S. Mondal ,&nbsp;M. Sen ,&nbsp;S.K. Makineni ,&nbsp;P. Ghosh ,&nbsp;A. Sarkar ,&nbsp;R. Kapoor ,&nbsp;S. Suwas","doi":"10.1016/j.jnucmat.2024.155487","DOIUrl":"10.1016/j.jnucmat.2024.155487","url":null,"abstract":"<div><div>In this investigation, the effect of 5.6 MeV proton irradiation on the microstructure and mechanical properties of coarse grained (CG) and nanocrystalline (NC) Nb-1wt.%Zr (NZ) has been analysed. Bulk nanocrystalline microstructure was obtained by subjecting the alloy to room temperature high pressure torsion under 6 GPa hydrostatic pressure and 5 rotations. The CG and NC samples were irradiated at doses of 1.9 × 10¹⁷ p/cm² and 1.8 × 10¹⁷ p/cm², respectively. Microstructural parameters like crystallite size, dislocation density, and dislocation arrangements were studied in detail using X-ray line profile analysis (XLPA) by Convolutional Multiple Whole Profile (CMWP) fitting. Microscopic observations were made with electron microscopy techniques in the scanning and transmission modes. Differential Scanning Calorimetry (DSC) was performed to estimate the concentration of vacancies after HPT processing and irradiation. Tensile tests of irradiated CG and NC irradiated samples were performed and compared to those in unirradiated conditions. In the NC condition, not only did the irradiated sample show higher ultimate tensile strength but also twice the amount of uniform elongation as compared to the irradiated CG sample. The fracture surface clearly exhibited this higher plasticity post-irradiation in the NC samples. The change in deformation mechanisms due to nano-structuring of the microstructure has been anticipated to be a reason for the increase in ductility in a single-phase alloy has been explained thereafter.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155487"},"PeriodicalIF":2.8,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652815","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Synthesis and characterization of multi-component borosilicate glass beads for radioactive liquid waste immobilisation","authors":"Sitendu Mandal ,&nbsp;Gattu Suneel ,&nbsp;Jayaprakasam Selvakumar ,&nbsp;Kaushik Biswas ,&nbsp;Srikrishna Manna ,&nbsp;Sourav Nag ,&nbsp;Balram Ambade","doi":"10.1016/j.jnucmat.2024.155485","DOIUrl":"10.1016/j.jnucmat.2024.155485","url":null,"abstract":"<div><div>High-level radioactive liquid waste (HLW) is immobilized in a glass matrix through a process called vitrification. In this process, HLW and glass-forming oxides are combined in a pre-determined ratio within a glass melter to produce a vitrified waste form. The properties of this waste form, including its ability to accommodate different radioactive isotopes, depend on the composition of the base glass.</div><div>In the present study, multi-component amorphous borosilicate-based glasses (SiO₂-B₂O₃-Na₂O-TiO₂-Fe₂O₃-CaO-K₂O) in bead form (diameter 2–3 mm) were developed. The elemental composition of the glass beads (GBs) was analyzed using an optical emission spectrometer. Additionally, the GBs underwent various physico-chemical analyses, including functional group identification, thermal, electrical, and mechanical properties, as well as viscosity and chemical durability assessments, to identify the optimal glass compositions. The influence of Na₂O on the pouring temperature was also examined. Crushing strength and attrition rate measurements were conducted to confirm the suitability of GBs for remote feeding into the melter. The GBs developed in the study are unique, with significant potential for worldwide use in vitrification facilities, particularly in continuous vitrification systems employing Joule Heated Ceramic Melter (JHCM) technology.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155485"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652878","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Length scale effects of micro- and meso‑scale tensile tests of unirradiated and irradiated Zircaloy-4 cladding","authors":"S. Lam ,&nbsp;D. Frazer ,&nbsp;F. Cappia ,&nbsp;M. Nelson ,&nbsp;S. Samuha ,&nbsp;S. Pitts ,&nbsp;B. Harris ,&nbsp;P. Hosemann","doi":"10.1016/j.jnucmat.2024.155469","DOIUrl":"10.1016/j.jnucmat.2024.155469","url":null,"abstract":"<div><div>Zircaloy-4 is an essential material for cladding structures within fission-based reactors. To explore the changes in properties measured on differing length scales, FIB-machined micro-scale tensile tests were performed on both irradiated and control groups of Zircaloy-4. This was correlated with tensile testing on femtosecond laser-machined meso‑scale specimens. Pronounced size effects were found when varying specimen geometry. Increases in tensile geometry size were associated with a reduction in measured yield stress for both irradiated and unirradiated samples. Meso-scale testing found strength and strain values similar to that of bulk-scale testing.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155469"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571579","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Effects of Ce addition on the morphology, crystal and metal/oxide interface structures of nanoparticles in FeCrAl-ODS steels","authors":"Tian-Xing Yang,&nbsp;Peng Dou,&nbsp;Chang-Jun Zhou","doi":"10.1016/j.jnucmat.2024.155484","DOIUrl":"10.1016/j.jnucmat.2024.155484","url":null,"abstract":"<div><div>FeCrAl oxide dispersion strengthened (ODS) steel is one of the most promising candidate cladding materials in generation IV nuclear reactors due to its exceptional macro-properties. To address the stringent performance requirements of supercritical water-cooled reactors (SCPWRs), two FeCrAl-ODS steels, i.e., 3Al–0.1Ti (Fe–16Cr–3Al–0.1Ti–0.34Y₂O₃) and 2Al–0.1Ti–0.35Ce (Fe–16Cr–2Al–0.1Ti–0.35Ce–0.36Y₂O₃), were developed. This study aims to investigate how Ce addition influences the microstructure and the formation mechanisms of various oxides in ODS steels. Therefore, the grain &amp; nanoparticle morphologies, and crystal &amp; interface structures of nano-scale oxides of the two ODS steels were studied by transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM) and high-resolution transmission electron microscopy (HRTEM). The mean grain diameter of 3Al–0.1Ti and 2Al–0.1Ti–0.35Ce is 1.1 μm and 0.82 μm, respectively. Compared with 3Al–0.1Ti, the average diameter of particles of 2Al–0.1Ti–0.35Ce is relatively smaller. The results indicate that adding Ce can refine the grains and nano-sized particles. For 3Al–0.1Ti, the main particles are Y–Al–O with a proportion of ∼81.4 %. For 2Al–0.1Ti–0.35Ce, the main particles are Y–Ce and Y–Ti oxides with quantity ratios of ∼52.2 % and ∼22.1 %, respectively, while the quantity ratio of Y–Al oxides is only 12.3 %. This indicates that adding Ce can impede the occurrence of Y–Al–O while facilitating the generation of Y–Ce–O. Moreover, it is the first time that Y₂Ce₂O₇ oxide has been detected in yttria-added ODS steels with Ce. The findings obtained from this study provide key insights into the mechanisms of oxide formation &amp; polymorphic transitions, and microstructural differences due to Ce addition. This will provide pivotal direction for the optimization of alloy compositions, promoting the innovation of ODS steels. Additionally, the feasibility analysis of the two ODS steels indicates their applicability to the SCPWR fuel cladding.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155484"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652828","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Investigation of radiation-induced segregation at fully characterized coherent twin boundaries in proton-irradiated 316L stainless steel","authors":"Sara K. Wonner,&nbsp;Pascal Bellon","doi":"10.1016/j.jnucmat.2024.155470","DOIUrl":"10.1016/j.jnucmat.2024.155470","url":null,"abstract":"<div><div>The effect of grain boundary character on radiation-induced segregation (RIS) is investigated in a 316L austenitic stainless steel irradiated with 2 MeV protons at 360 °C. Orientation imaging microscopy is employed to select specific grain boundaries (GBs), including Σ3{111} coherent twin boundaries, fully characterized by a five-degree of freedom analysis, as well as high angle GBs. Chemical maps along these GBs below the irradiated surface, at depths corresponding to damage levels ranging from 2.3 dpa to 4.2 dpa, are acquired using energy-dispersive spectrometry in a scanning transmission electron microscope (STEM-EDS). RIS levels are defined as elemental GB excess quantities and are used to compare RIS at twin boundaries and high-angle GBs. These measurements are complemented by the analysis of void distributions near GBs and by characterizing the structure of coherent twin GBs prior to and after irradiation using high-resolution STEM imaging. In light of the results obtained in this work, the evolution of the efficiency for point defect elimination of coherent twin GBs with the irradiation dose is discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155470"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142705990","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Post-irradiation examination of UN-Mo-W fuels for space nuclear propulsion","authors":"Sarah A. Khan ,&nbsp;Jason L. Schulthess ,&nbsp;Indrajit Charit ,&nbsp;Aaron Craft ,&nbsp;William Chuirazzi ,&nbsp;Jatuporn Burns ,&nbsp;David Frazer ,&nbsp;Nicolas Woolstenhulme ,&nbsp;Robert O'Brien","doi":"10.1016/j.jnucmat.2024.155476","DOIUrl":"10.1016/j.jnucmat.2024.155476","url":null,"abstract":"<div><div>The National Aeronautics and Space Administration's return to space nuclear propulsion stems from the need for a more efficient method of space travel. Nuclear thermal propulsion systems have been shown to be two times more efficient than chemical propulsion. NASA's Sirius program was created to fabricate and test fuels for space nuclear propulsion, specifically to determine their performance under prototypical startup conditions. The Sirius project featured 4 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A, -2B, and -3 featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). This study discusses the Sirius-2A and -2B irradiation experiments at the Idaho National Laboratory, specifically their performance under irradiation at the Transient Reactor Test Facility. It was found that the fuel samples overall did not exhibit significant cracking, though the Sirius-2A fuel did have one large crack on the surface of the fuel. There was minimal hydrogen absorption in the samples, though it is unknown if the absorption occurred during irradiation or during fabrication. Mechanical testing indicated that the UN fuel demonstrated ceramic behavior as expected, and the Mo/W matrix demonstrated linear elastic behavior to failure.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155476"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552505","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Atomistic mechanism of activation controlled liquid metal corrosion at the Fe-Pb interface","authors":"Ilia V. Voronov ,&nbsp;Vladislav S. Nikolaev ,&nbsp;Alexey V. Timofeev ,&nbsp;Vladimir V. Stegailov","doi":"10.1016/j.jnucmat.2024.155483","DOIUrl":"10.1016/j.jnucmat.2024.155483","url":null,"abstract":"<div><div>A bcc iron bicrystal in contact with liquid lead is studied in molecular dynamics simulations to describe the atomistic mechanism of liquid metal corrosion in the activation controlled case. In this process, the main structural features involved are Fe grain boundaries and Fe-Pb interfaces. The atomistic model considered reveals that the interplay of atomic processes such as surface self-diffusion of Fe and dissolution of Fe into Pb determines the mechanism and kinetics law of liquid metal corrosion. Analysis of the proposed mechanism explains the dependence between the kinetics of liquid metal corrosion and the grain size of the specimen.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155483"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142587028","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Bulk, overlap and surface effects of swift heavy ions in CeO2","authors":"R.A. Rymzhanov ,&nbsp;A.E. Volkov ,&nbsp;V.A. Skuratov","doi":"10.1016/j.jnucmat.2024.155480","DOIUrl":"10.1016/j.jnucmat.2024.155480","url":null,"abstract":"<div><div>Formation of tracks of swift heavy ions decelerating in the electronic stopping regime in CeO₂ was studied, combining the Monte Carlo code TREKIS with molecular dynamics. We show that strong lattice disordering (melting) followed by structure recovery form finally a damaged ion track consisting of a discontinuous crystalline region in CeO₂. Normal ion impacts result in appearance of spherical crystalline hillocks on CeO₂ surface. The solid-vacuum interface strongly suppresses the recrystallization of the near-surface layers, forming conically shaped tracks with several tens of nanometers lengths. Grazing ion irradiation induces intensive material expulsion from the surface forming finally grooves surrounded by nanohillocks. The processes of surface nanostructures formation is similar to those observed previously in CaF₂ which has the similar crystalline structure, however requires much longer recrystallization time. Recent experimental data confirm the simulation results.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155480"},"PeriodicalIF":2.8,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571580","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}