Journal of Nuclear Materials最新文献

{"title":"Transmission Electron Microscopy Characterization of Fuel Cladding Chemical Interaction between Minor Actinides bearing U-Pu-Zr Fuel and AIM1 Cladding","authors":"Di Chen,&nbsp;Jatuporn Burns,&nbsp;Karen E. Wright,&nbsp;Daniele Salvato,&nbsp;Tiankai Yao,&nbsp;Luca Capriotti","doi":"10.1016/j.jnucmat.2025.155667","DOIUrl":"10.1016/j.jnucmat.2025.155667","url":null,"abstract":"<div><div>Minor actinides (MA) significantly contribute to the long-term radiotoxicity of spent nuclear fuel (SNF). Separating MA from SNF and incorporating it into metallic fuels for fast reactor transmutation is a potential method to reduce this radiotoxicity. This study focuses on transmission electron microscopy characterization of two samples from the fuel cladding chemical interaction (FCCI) region of an americium (Am) and neptunium (Np)-bearing (MA-bearing) uranium-plutonium-zirconium (U-Pu-Zr) fuel irradiated in the Phenix fast reactor to 9.5 % FIMA burnup at approximately 550 °C cladding temperature. The results show that despite the complex chemical interactions between MA and AIM1 cladding elements, excessive FCCI was not induced, and Am penetration depth in the cladding limited to less than 4 μm. Np remained mostly inside fuel. The Zr-rich compounds layer effectively limited the accumulation of lanthanide on the inner cladding surface. Overall, the FCCI behavior between investigated MA-bearing U-Pu-Zr fuel and AIM1 cladding is benign.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155667"},"PeriodicalIF":2.8,"publicationDate":"2025-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454719","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 vacancy dynamic evolution model based on rate theory and its influence on hydrogen isotope retention behavior","authors":"Fei Sun ,&nbsp;Xiangming Lin ,&nbsp;Yijin Huang ,&nbsp;Jipeng Zhu ,&nbsp;Long Cheng ,&nbsp;Yifan Zhang ,&nbsp;Hai-Shan Zhou ,&nbsp;Lai-Ma Luo ,&nbsp;Yasuhisa Oya ,&nbsp;Yucheng Wu","doi":"10.1016/j.jnucmat.2025.155676","DOIUrl":"10.1016/j.jnucmat.2025.155676","url":null,"abstract":"<div><div>In fusion environments, plasma-facing materials are subjected to the bombardment by high-energy neutrons and large fluxes plasmas, resulting in the formation of numerous irradiation defects. These defects act as trapping sites for fuel hydrogen isotopes, increasing the tritium retention. Under thermal loads at high temperatures, these defects undergo dynamic changes, leading to complex effects on hydrogen isotope transport behaviors. Understanding the evolution of defects at different temperatures and elucidating their impact on hydrogen isotope retention is essential. In this work, a dynamic evolution model of typical vacancy defects and their clusters was constructed based on rate theory and integrated into the hydrogen isotope transport model. The simulation results were compared with experimental data, showing a good agreement between the two. Further results indicated that increasing the annealing temperature reduces defect concentrations. At high temperatures, large vacancy clusters dissociate into smaller ones, with some smaller clusters annihilating and others associating to form larger clusters. In addition, the processes of vacancy clusters association and dissociation dominate this evolution. Regarding hydrogen isotope retention, defect evolution affects the concentration of deuterium (D) atoms within the material. After D is introduced into tungsten, the concentration of D corresponds to the concentration of vacancy clusters. It is found that during thermal desorption, the desorption peak corresponding to the largest vacancy clusters is more pronounced, while other clusters show less significant peaks. This is because some of the D atoms de-trapped from the vacancy clusters are released to the material surface, while others are re-trapped by larger vacancy clusters. The findings of this work will contribute to a more accurate assessment of hydrogen isotope retention under defect dynamic evolution in future fusion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155676"},"PeriodicalIF":2.8,"publicationDate":"2025-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143360821","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":"Atomic-scale mechanisms of He-induced dislocation loop growth in α-Fe from molecular dynamics simulations","authors":"Ziqiang Wang,&nbsp;Xiangyan Li,&nbsp;Yange Zhang,&nbsp;Yichun Xu,&nbsp;Changsong Liu,&nbsp;Xuebang Wu","doi":"10.1016/j.jnucmat.2025.155672","DOIUrl":"10.1016/j.jnucmat.2025.155672","url":null,"abstract":"<div><div>The interaction between helium (He) and irradiation damage is crucial for understanding the degradation mechanisms and performance of structural materials for fusion applications. However, the atomic-scale mechanisms underlying the helium influence on dislocation loop growth remain poorly understood. In this work, we investigated the interactions between He and interstitial dislocation loops in bcc iron using molecular dynamics simulations. It was found that the formation of He bubbles around dislocation loops enhances loop growth through the dislocation climb mechanism. An increased He implantation rate results in a higher growth rate of dislocation loops, with more He bubble nucleation on the loops leading to greater loop extension. Furthermore, the interactions between dislocation loops and immobile He bubbles contribute to loop enlargement through the dislocation loop absorption mechanism. Larger He bubbles induce a more substantial rise in loop length. These findings provide valuable insights into the atomic-scale processes by which He promotes dislocation loop growth in α-Fe, advancing our understanding of He-induced irradiation damage in fusion materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155672"},"PeriodicalIF":2.8,"publicationDate":"2025-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143237314","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":"Assessment of structural stability and leaching characteristics of phosphate-based geopolymer waste form containing radioactive spent ion exchange resins","authors":"Byoungkwan Kim ,&nbsp;Younglim Shin ,&nbsp;Solmoi Park ,&nbsp;Brant Walkley ,&nbsp;Seokyoung Oh ,&nbsp;Jaehyuk Kang ,&nbsp;Wooyong Um","doi":"10.1016/j.jnucmat.2025.155671","DOIUrl":"10.1016/j.jnucmat.2025.155671","url":null,"abstract":"<div><div>The use of hydrated waste forms, such as cement, for immobilizing radioactive spent ion-exchange resins (IERs) is unsuitable due to low waste loading, high leaching of radionuclides, and poor durability. Here, simulant spent IERs were immobilized by a phosphate-based geopolymer (P-GP) for the first time. The 7-day compressive strength of the P-GP waste form was inversely proportional to waste loading because the pore size of the P-GP waste form increased with increasing waste loading. The P-GP waste forms with 40 wt% spent IERs satisfied all of South Korea's waste acceptance criteria for compressive strength, thermal cycling, water immersion, and gamma irradiation tests. The leaching behaviors of Co, Cs, and Sr differed from those of alkali-activated materials, but the leaching index exceeded the criterion value of 6.0. The leaching mechanism was governed by the combination of surface wash-off and diffusion or solely diffusion. The P-GP waste form could play as a primary physical barrier against releasing radionuclides. In addition, the geopolymer waste form did not undergo significant structural changes after waste acceptance criteria tests, indicating that it can efficiently immobilize spent IERs. Our findings can contribute new insights into efficient waste form materials for immobilizing radioactive spent IERs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155671"},"PeriodicalIF":2.8,"publicationDate":"2025-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143237313","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":"Exploring the influence of alloying elements (Zr, Si, and V) on the corrosion behavior of FeCoNiCr-based high entropy alloys in supercritical water environment","authors":"Yuxuan Xia ,&nbsp;Shanliang Zhou ,&nbsp;Zihang Liang ,&nbsp;Xi Huang ,&nbsp;Qi Zhao ,&nbsp;Huiwang Huang ,&nbsp;Haidan Zhao ,&nbsp;Yilei Zhan ,&nbsp;Changfu Wang ,&nbsp;Yujun Xie","doi":"10.1016/j.jnucmat.2024.155605","DOIUrl":"10.1016/j.jnucmat.2024.155605","url":null,"abstract":"<div><div>This study investigated the effect of alloying elements Zr, V, and Si on the corrosion behavior of FeCoNiCr-based high entropy alloys in supercritical water. Adding Zr decreased corrosion resistance owing to the formation of a thinner Cr₂O₃ outer layer and a mixed Cr₂O₃ and ZrO₂ inner layer. Adding Si formed a complex oxide film with a triple-layer structure with pores and cracks, accelerating corrosion. By contrast, adding V improved corrosion resistance because the formation of a compacted V_0.87Cr_0.13O_2.7 layer inhibited the diffusion of anions and cations. The corrosion mechanisms were also discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155605"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170535","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":"Comparative analysis of the long-term strength of Russian ferritic-martensitic reactor steels","authors":"N.V. Kataeva ,&nbsp;V.V. Sagaradze ,&nbsp;V.A. Zavalishin ,&nbsp;K.A. Kozlov ,&nbsp;V.A. Sirosh ,&nbsp;M.V. Leont'eva-Smirnova ,&nbsp;A.A. Nikitina","doi":"10.1016/j.jnucmat.2024.155575","DOIUrl":"10.1016/j.jnucmat.2024.155575","url":null,"abstract":"<div><div>The paper presents the results of long-term high-temperature creep tests of Russian reactor steels with ferritic-martensitic structure (the duration of some measurements exceeded 8 years). In the current study, the structural-phase transformations, characteristics of creep and long-term strength at 650 °C, 670 °C, and 700 °C under 60–140 MPa in oxide-free and oxide containing steels were determined. The creep tests were performed on specially designed transverse micro-specimens prepared from fuel elements cladding used in the fast-neutron reactor. The creep velocity of the ferritic-martensitic reactor steels was established to be specified by resistance of lath martensite and ferrite structures to diffusion processes of return and recrystallization. The most heat-resistant oxide-free steel contains the largest amount of refractory elements and carbides. The best heat resistance was observed for the steel hardened with thermal-resistant yttrium-titanium nanooxides. The samples made of this steel demonstrated one order less creep velocity at 700 °C under 100 MPa and 100-fold time to fracture in comparison with the oxide-free reactor steels.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155575"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170554","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":"Chemical characterization of uniform and localized oxides formed on Zr-2.5 (wt. %) Nb alloy","authors":"Akanksha Samanta ,&nbsp;Suparna Banerjee ,&nbsp;Rekha Rao ,&nbsp;V. Srikanth ,&nbsp;Y. Sunitha ,&nbsp;Santu Kaity ,&nbsp;J.V. Ramana ,&nbsp;A.K. Tyagi","doi":"10.1016/j.jnucmat.2024.155546","DOIUrl":"10.1016/j.jnucmat.2024.155546","url":null,"abstract":"<div><div>Uniform and localized oxide morphologies formed on Zr-2.5(wt.%) Nb alloy in water and steam environments have been studied using electron microscopy techniques such as SEM and EPMA. The uniform oxidation was characterized by a compact and thin oxide layer. In contrast, the nodules showed localized thicker oxide growth with cracked and porous morphology. Significant segregation of Nb at the surface of the oxide nodule was observed, leading to destabilization and spallation of the oxide. The chemical states of Zr, Nb, and O in the two types of oxides were analyzed by XPS. The atomic composition of oxides in terms of O and M fractions was evaluated by p-EBS. The study revealed the oxide nodule to be richer in oxygen in comparison to the surrounding uniform oxide. The phase characteristics of these oxides were studied by Raman spectroscopy. The effect of initial surface in-homogeneities on nodule nucleation has been discussed. Polishing made the surface free of initial in-homogeneities and resulted in uniform oxidation on the surface.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155546"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170558","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":"Fracture behavior of irradiation induced nanocrystalline UO2 studied by in-situ mechanical testing in transmission electron microscopy","authors":"Laura Hawkins ,&nbsp;Ana Garcia Caraveo ,&nbsp;David Frazer ,&nbsp;Fabiola Cappia ,&nbsp;Tianyi Chen ,&nbsp;Collin Knight ,&nbsp;Jeffrey J. Giglio ,&nbsp;Tiankai Yao","doi":"10.1016/j.jnucmat.2024.155571","DOIUrl":"10.1016/j.jnucmat.2024.155571","url":null,"abstract":"<div><div>Uranium Dioxide (UO₂) is widely used as a fuel in current light water reactors (LWRs). Upon accumulation of radiation damage, LWR UO₂ fuel pellets start to develop a different microstructure at the pellet periphery when fuel burnup exceeds 45–50 GWd/tHM. The resulting porous, nanocrystalline microstructure is one of the most prominent microstructural changes occurring in such fuel. Its fracture mechanisms, which causes fuel fine fragmentation, could impact safety limits when the cladding breaches. Direct measurements of these properties are challenging, therefore a surrogate obtained via ion irradiation can be used. In this study, multiple microcantilevers were fabricated by focused ion beam from both fresh UO₂ and UO₂ irradiated with 84 MeV Xe²⁶⁺ ions to a peak dose of 1357 displacements per atom (dpa). The irradiation produced a pseudo high burnup structure approximately 2 µm below the surface. In-situ nano-mechanical bending tests were conducted to investigate the fracture behavior and the effect of the surrogate UO₂ high burnup structure on local fracture properties. Fresh UO₂ fuel was observed to fracture in transgranular mode without nucleation or movement of dislocations. However, the Xe-irradiated nanocrystalline microcantilevers fractured along the grain boundaries, with no influence from the pre-existing micro-cracks in the microcantilever. Fracture toughness for this type of surrogate high burnup UO₂ structure is reported for the first time in literature. Both the fracture stress and toughness show degradation for UO₂ as a result of Xe-irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155571"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170561","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":"Self-ion-irradiation-induced grain formation in nickel","authors":"S.E. Donnelly ,&nbsp;G. Greaves ,&nbsp;F. Granberg ,&nbsp;J. Sharp ,&nbsp;A.H. Milston ,&nbsp;J.A. Hinks ,&nbsp;K. Nordlund","doi":"10.1016/j.jnucmat.2024.155588","DOIUrl":"10.1016/j.jnucmat.2024.155588","url":null,"abstract":"<div><div>A series of experiments has been conducted in which thin foils containing large polycrystals of Ni (single crystals from the perspective of transmission electron microscopy) have been irradiated with 300 keV Ni ions at temperatures from 25° to 475°C. The aim was to examine the fundamental aspects of the build-up of extended defects in a “simple” system with no implantation of foreign species and without the likelihood of segregation, precipitation or formation of new phases. Experiments were carried out using the MIAMI-2 facility in which the development of radiation damage is observed (and recorded) whilst ion-irradiating in-situ in a transmission electron microscope. Surprisingly, all irradiations of the electrochemically-thinned foils of Ni resulted in the accumulation of dislocations to form low-angle grain boundaries such that single crystal material was converted into a series of grains, each typically less than 200 nm in width but generally more than 1 <em>µ</em>m in length with the long axis approximately parallel to the edge of the foil. The early stages of this process have been modelled using Molecular Dynamics simulations and an interpretation of this process of radiation-induced grain-boundary formation is discussed in terms of the coupled effects of irradiation, temperature and stress induced by the radiation damage. The stress arises due to swelling in the thin irradiated region of the jet-polished specimens (with a wedge-shaped radial cross section) which is constrained by deeper-lying unirradiated material. The position in which a grain boundary forms is determined by the interaction of glisssile dislocations with the stress induced by the radiation-damaged layer and that from a neighbouring boundary.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155588"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170986","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":"Dissolution behavior of U3O8, FeUO4, and UO2-Zr-stainless steel system samples generated in an oxidative atmosphere in the presence of malonic acid","authors":"Ryutaro Tonna ,&nbsp;Takayuki Sasaki ,&nbsp;Yoshihiro Okamoto ,&nbsp;Taishi Kobayashi","doi":"10.1016/j.jnucmat.2024.155561","DOIUrl":"10.1016/j.jnucmat.2024.155561","url":null,"abstract":"<div><div>Understanding the dissolution behavior of the UO₂-Zr-stainless steel (SS) system in the presence of organic acids is crucial for the safe decommissioning of the Fukushima Daiichi Nuclear Power Plants, as well as for the future treatment toward deep geological disposal of fuel debris. The dissolution behavior of UO₂-Zr-SS samples heated under oxidative conditions, along with the single uranium phases formed in the samples (U₃O₈ and FeUO₄)_, was investigated through static leaching tests using malonic acid. Malonic acid promoted the dissolution of both U₃O₈ and FeUO₄ solid phases owing to complex formation. The uranium concentrations of U₃O₈ and FeUO₄ increased with malonic acid concentration and matched at steady state. In the absence of malonic acid, the U concentration of FeUO₄ was less than its solubility because of uranium adsorption on an iron hydrolysis solid phase. The uranium concentration after long immersion of the UO₂-Zr-SS system samples could be explained by the behavior of U₃O₈ and FeUO₄. In the absence of malonic acid, U concentration decreased owing to adsorption onto the iron solid phase, similar to the behavior of FeUO₄. In contrast, in the presence of malonic acid, U concentration was consistent with that observed for U₃O₈ and FeUO₄.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155561"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171602","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}