Journal of Nuclear Materials最新文献

{"title":"Development of an advanced hydride reorientation model for Zircaloy cladding and its experimental validation","authors":"Changhyun Jo,&nbsp;Dahyeon Woo,&nbsp;Youho Lee","doi":"10.1016/j.jnucmat.2024.155445","DOIUrl":"10.1016/j.jnucmat.2024.155445","url":null,"abstract":"<div><div>Hydride reorientation, which occurs under hoop stress during cooling, stands out as a primary mechanism for material degradation in spent fuel management. The radial hydride fraction (RHF) is strongly involved in the mechanical integrity of cladding, highlighting the necessity for a robust modeling framework for quantitative analysis. However, the predictability of previous thermodynamic models for hydride reorientation in reactor-grade Cold Worked Stress Relieved (CWSR) Zircaloy has been hindered due to the intricate nature of hydride reorientation and the difficulties in characterizing microstructures. Recent successful EBSD characterization of reactor-grade CWSR Zircaloy has revealed valuable insights into microstructural characteristics of hydrides, enabling advancements in the modeling framework of hydride reorientation. This study aims to develop a thermodynamic model specifically focused on predicting the RHF. The developed thermodynamic model, based on classical nucleation theory, integrates aforementioned microstructural findings, combined with the Hydride-Nucleation-Growth-Dissolution (HNGD) model to capture transient precipitation behavior during cooling. Extensive experimental validations demonstrate enhanced predictability of the model. Additionally, the study examines the sensitivities of hydride reorientation to hydrogen concentration, applied stress, and cooling rate. It also provides predictions on reorientation behavior for engineering implications such as extension of wet storage, matrix hardening, recrystallization, and thermal cycling, supported by plausible explanations rooted in the underlying physical mechanisms elucidated through the model.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155445"},"PeriodicalIF":2.8,"publicationDate":"2024-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661915","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":"Displacement cascade bombardment of delta-hydrides in alpha-zirconium","authors":"Jose F. March-Rico,&nbsp;Richard W. Smith,&nbsp;Brendan M. Ensor","doi":"10.1016/j.jnucmat.2024.155446","DOIUrl":"10.1016/j.jnucmat.2024.155446","url":null,"abstract":"<div><div>One of the factors affecting the in-pile performance of Zr-based alloys is the precipitation of hydrides once H concentrations exceed the terminal solubility limit. H transport and hydride precipitation/dissolution is commonly modeled in codes such as BISON, but most of the experimental data supporting these models has been collected on unirradiated materials. As such, there is considerable uncertainty as to the influence of irradiation effects. In this work, molecular dynamics simulations of displacement cascades were performed on δ-hydrides to elucidate: 1) the extent of H dissolution following cascade impacts and 2) any alterations to defect production characteristics when compared to cascades in bulk Zr. The immediate amount of H dissolved in a high-energy cascade impact is notable, but a considerable fraction of the dissolved H atoms are rapidly re-absorbed into the hydride at reactor-relevant temperatures. The amount of dissolved H also decreases with increasing hydride size. When considering the expected volume fractions of hydrides, it is not expected that the irradiation-induced H dissolution rate will significantly affect the availability of H in the Zr lattice. In terms of defect production, cascades which overlap δ-hydrides produced an order of magnitude more stable defects than equivalent-energy cascades in bulk Zr. Vacancy defects are predominantly contained within the hydride structure while interstitials clusters are found adjacent to the hydride surface. Interstitials are strongly repelled by the hydride structure which may drive the expulsion of cascade-generated interstitials to the hydride surface and impede athermal recombination. Thus, the interatomic potential used in this work predicted a significant alteration to the defect survival efficiency and a stark production bias in the availability of mobile defects in bulk Zr following hydride-overlapped displacement cascades.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155446"},"PeriodicalIF":2.8,"publicationDate":"2024-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537324","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":"High-temperature structure, elasticity, and thermal expansion of ε-ZrH1.8","authors":"James R. Torres ,&nbsp;Christopher A. Mizzi ,&nbsp;Daniel A. Rehn ,&nbsp;Tyler Smith ,&nbsp;Scarlett Widgeon Paisner ,&nbsp;Adrien J. Terricabras ,&nbsp;Darren M. Parkison ,&nbsp;Sven C. Vogel ,&nbsp;Caitlin A. Kohnert ,&nbsp;Mathew L. Hayne ,&nbsp;Thomas J. Nizolek ,&nbsp;M.A. Torrez ,&nbsp;Tannor T.J. Munroe ,&nbsp;Boris Maiorov ,&nbsp;Tarik A. Saleh ,&nbsp;Aditya P. Shivprasad","doi":"10.1016/j.jnucmat.2024.155437","DOIUrl":"10.1016/j.jnucmat.2024.155437","url":null,"abstract":"<div><div>Zirconium hydride is a promising candidate material for nuclear microreactor applications as a solid-state moderator component, owing to its favorable neutronics properties and good thermal stability over other metal hydrides. In the present work, the crystal structure, thermal expansion, and elastic properties of the hydrogen-rich ε phase hydride were measured at elevated temperatures in the range 300–900 K. Samples were prepared by direct hydriding Zircaloy-4 metal – a nuclear-grade zirconium alloy. Room-temperature lattice parameters agree well with those reported from literature for unalloyed zirconium hydride and fall within an observed quadratic H-content dependence. The coefficients of thermal expansion, determined from lattice expansion and dilatometry, agree well within our work but were about 30 % lower than those reported by others for unalloyed hydrides. Density functional theory-based molecular dynamics simulations were used to compare with thermal expansion and elasticity measurements. Results showed lattice parameter temperature dependence and slope of thermal expansion align with those from measurements. Based on diffraction scans at select temperatures, ε phase remained stable in air up to at least 770 K. Likewise, dilatometry showed smooth thermal expansion up to the thermal decomposition temperature around 950 K. The precise decomposition temperature was not determined via diffraction due to sparse scanning. The complete elastic property measurements were gathered for ε-phase Ziracloy-4 hydride for the first time. Young's modulus was lower compared to the metal and δ hydride phases. High-temperature elasticity measurements were limited to &lt;350 K due to acoustic dissipation effects.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155437"},"PeriodicalIF":2.8,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432545","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":"On the theory of the nucleation of gas bubbles at grain boundaries and incoherent inclusions","authors":"M.S. Veshchunov","doi":"10.1016/j.jnucmat.2024.155443","DOIUrl":"10.1016/j.jnucmat.2024.155443","url":null,"abstract":"<div><div>On the base of the critical analysis of two-dimensional models of the nucleation of gas filled bubbles at grain boundaries of helium-implanted specimens under the action of tensile stresses, a new model is developed within the framework of the Reiss theory of homogeneous nucleation in binary systems. This approach considers that gas bubbles are formed as a result of agglomeration in a binary system of vacancies and gas atoms at grain boundaries, avoiding significant simplifications of previous models based on the classical nucleation theory for single-component (unary) systems. The new model is extended to consider the nucleation of Xe bubbles at grain boundaries in UO₂ under irradiation conditions and can be used for numerical analysis of experimental observations after the foreseen implementation in a fuel performance code. A similar approach can be applied to the nucleation and growth of gas bubbles on incoherent inclusions, such as those observed in irradiated ODS steels.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155443"},"PeriodicalIF":2.8,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432546","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":"Microstructural changes and irradiation hardening behavior of V-4Cr-4Ti alloys irradiated with He ions using flash-electropolishing","authors":"Ken-ichi Fukumoto ,&nbsp;Yichen Zou ,&nbsp;Takuya Nagasaka ,&nbsp;Ryoya Ishigami","doi":"10.1016/j.jnucmat.2024.155438","DOIUrl":"10.1016/j.jnucmat.2024.155438","url":null,"abstract":"<div><div>The flash-electropolishing of focused ion beam samples for V-4Cr-4Ti alloys is established, and the microstructures of high-purity V-4Cr-4Ti alloys after He ion irradiation are examined by transmission electron microscopy from room temperature to 700 °C. The correlation between irradiation hardening behavior and microstructural changes is clarified. During room temperature irradiation, defect clusters are formed at shallow positions in the specimens and no He bubbles are observed at the damage peak position. In contrast, 500 and 700 °C, TiCON precipitates are predominantly formed and He bubbles and voids were formed at the damage peak position. The results of nanoindentation tests and a comparison of irradiation hardening by irradiation damage indicate that the obstacle barrier strength factorαof TiCON is 0.45 while that of the irradiation defect clusters irradiated at room temperature is 0.10. Irradiation damage in the He ion range extends toward the interior of the specimens with increasing irradiation temperature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155438"},"PeriodicalIF":2.8,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432462","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":"The effect of deuterium on defect production in irradiated tungsten","authors":"V. Lindblad ,&nbsp;D.R. Mason ,&nbsp;F. Granberg","doi":"10.1016/j.jnucmat.2024.155422","DOIUrl":"10.1016/j.jnucmat.2024.155422","url":null,"abstract":"<div><div>For fusion test reactors and power plants, one significant concern is the retention of hydrogen isotopes in the wall materials. The build-up of the radioactive and scarce fuel isotope tritium is of special concern, but knowing the retention of the other isotopes, such as deuterium, is also important. Deuterium is known to affect the mechanical properties of the wall material and most experiments are carried out on deuterium retention as it is safer to use than tritium. In addition to affecting the mechanical properties of the wall material, deuterium retention has been observed to affect the defect accumulation in the material. In this study, we investigate the phenomena and mechanisms responsible for the greater defect accumulation observed in tungsten when deuterium is present during irradiation. This is achieved computationally, utilizing molecular dynamics simulations and appropriate analysis tools. We found that deuterium will affect both the primary defect production as well as the recombination rate of defects in irradiated tungsten.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155422"},"PeriodicalIF":2.8,"publicationDate":"2024-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432569","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Crevice corrosion behavior of 316L austenitic steel in static liquid lead-bismuth eutectic at 550°C","authors":"Yuji Huang ,&nbsp;Fanqiang Meng ,&nbsp;Jiajian Shi ,&nbsp;Lijun Zhang ,&nbsp;Rui Yuan ,&nbsp;Yingxue Chen ,&nbsp;Feifei Zhang","doi":"10.1016/j.jnucmat.2024.155441","DOIUrl":"10.1016/j.jnucmat.2024.155441","url":null,"abstract":"<div><div>Crevice corrosion is a significant form of localized corrosion that can lead to failures in structural components. However, there is a lack of research on this phenomenon in liquid lead-bismuth eutectic (LBE). In this study, the crevice corrosion behavior of 316 L exposed to stagnant LBE with the oxygen concentration of 1 × 10^–6 wt.% at 550 °C for 500 h was examined for the first time. Microstructural characterizations indicated 316 L is susceptible to crevice corrosion and reducing the crevices size facilitates a transition from oxidation to dissolution corrosion. Grain boundaries are able to provide more diffusion channels for oxygen, thereby promoting the development of Cr-rich oxides and chemical segregations beneath the surface scale. A simplified model elucidating the transportation of dissolved oxygen within the crevice environment of LBE has been developed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155441"},"PeriodicalIF":2.8,"publicationDate":"2024-10-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432460","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":"A finite element study on the irradiation-induced mechanical behaviors of aluminum-matrix radiation-shielding composites","authors":"Jiaqing Shi ,&nbsp;Zheng Lv ,&nbsp;Jian Wang ,&nbsp;Wentao Tang ,&nbsp;Yufei Liu ,&nbsp;Zenglin Yang ,&nbsp;Jian Yang ,&nbsp;Zhimin Yang ,&nbsp;Shuwang Ma","doi":"10.1016/j.jnucmat.2024.155440","DOIUrl":"10.1016/j.jnucmat.2024.155440","url":null,"abstract":"<div><div>Aluminum-matrix radiation-shielding composites play a crucial role in advanced nuclear energy systems and fuel containers owing to their shielding design flexibility and desired structural compatibility. After being irradiated by neutrons, the shielding composites undergo irradiation damage and exhibit irradiation-induced mechanical effects such as irradiation hardening and embrittlement, which directly threaten the industrial application of the material. In this study, a finite element method was used to investigate the irradiation-induced mechanical behavior of radiation-shielding B₄C_P-W_P/Al composites. Using published data on the post-irradiation mechanical property evolutions of the matrix and shielding particles, and incorporating mechanisms of irradiation hardening and embrittlement, a finite element model was developed to describe the deformation of pristine and post-irradiation composites. Simulations of the post-irradiation mechanical properties of the aluminum-matrix radiation-shielding composites were conducted. The simulation results successfully reproduced the experimental findings for both the Al matrix and composites after irradiation. Furthermore, the stress-strain responses and deformation behaviors of the composites at different stages of irradiation damage are discussed. Finally, based on the simulation results, an artificial neural network was trained to efficiently predict the irradiation-induced mechanical behavior of the composites.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155440"},"PeriodicalIF":2.8,"publicationDate":"2024-10-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432341","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":"TEM-EELS analysis reveals the W-atom mediated radiation-induced amorphization in M23C6","authors":"Sho Kano ,&nbsp;Huilong Yang ,&nbsp;Masami Ando ,&nbsp;Dai Hamaguchi ,&nbsp;Takashi Nozawa ,&nbsp;Hiroyasu Tanigawa ,&nbsp;Kenta Yoshida ,&nbsp;Tamaki Shibayama ,&nbsp;Hiroaki Abe","doi":"10.1016/j.jnucmat.2024.155439","DOIUrl":"10.1016/j.jnucmat.2024.155439","url":null,"abstract":"<div><div>To gain a mechanistic understanding of the phase stability of M₂₃C₆ upon irradiation, the bulk W-doped M₂₃C₆ (Cr-W-C system) in the range of 0–12 at.% W concentration was prepared and subjected to helium beam irradiation, following with a thorough electron energy loss spectroscopy (EELS) analysis. Radiation-induced amorphization (RIA) was observed only at the 4 W sample with a W concentration of ∼12 at.%. Analysis of the low-loss spectrum showed that the inelastic mean free path (<em>λ</em>) could be applied an effective indicator of the presence of an amorphous phase. The white line ratio of the carbon <em>K</em>-edge spectrum showed that the chemical bonding state in the crystalline state is mainly 2<em>p^3/2</em> bonding, and it changes to dominantly 2<em>p^1/2</em> bonding accompanying with the crystal-to-amorphous (c-a) transition. Discussion on the relationship between the change in <em>λ</em> (<em>Δλ</em>) and the lattice parameter (<em>Δa</em>) due to irradiation reveals that <em>Δa</em> is not dependent on <em>Δλ</em>, indicating that <em>Δλ</em> is mainly caused by the volume expansion due to the c-a transition. In addition, a crystalline state is remained even after a lattice parameter change of ∼1.5 % in 0 W and 1W-samples, whereas, a lattice expansion of ∼0.2 % would trigger the occurrence of crystal-to-amorphous transition in the 4W-sample. The detailed EELS analysis demonstrated that the constitutional W atoms play an important role in facilitating the occurrence of RIA in M₂₃C₆, that is, the phase instability accompanying the lattice expansion due to irradiation was emphasized by the addition of W in M₂₃C₆. The insights obtained here suggest that a higher W concentration in M₂₃C₆ is more susceptible to RIA, and therefore the resistance to amorphization is achievable by decreasing the W concentration in the steels.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155439"},"PeriodicalIF":2.8,"publicationDate":"2024-10-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432459","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":"Influence of liquid lead and lead-bismuth eutectic on three alumina forming austenitic (AFA) steels through slow strain rate testing","authors":"Christopher Petersson ,&nbsp;Peter Szakalos ,&nbsp;Rachel Pettersson ,&nbsp;Mats Lundberg","doi":"10.1016/j.jnucmat.2024.155415","DOIUrl":"10.1016/j.jnucmat.2024.155415","url":null,"abstract":"<div><div>Liquid metal embrittlement (LME) in three newly developed alumina-forming austenitic (AFA) alloys, two 50 kg batches and one 5-ton heat, was studied in the temperature range 350–600 °C in liquid Pb and 140–600 °C in LBE using slow strain rate testing (SSRT) in a low-oxygen environment. No significant decrease in the engineering strain was observed in either environment. However, the presence of secondary cracks along the length of the specimen and brittle intergranular areas on the fracture surfaces indicates that the AFA alloys do show a minor degree of embrittlement above 570 °C. This appears to be related to grain boundary wetting by Pb/LBE. At temperatures below 570 °C, this wetting effect does not seem to be strong enough to induce LME in the alloys, and their ability to form a sufficiently protective oxide means that they remain unaffected by LME. The results indicate that the AFA alloy group can perform sufficiently well in liquid Pb/LBE environments, and long-term testing should be carried out to determine their viability as candidate materials for use in Pb- and LBE-based cooling systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155415"},"PeriodicalIF":2.8,"publicationDate":"2024-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432451","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}