Journal of Nuclear Materials最新文献

{"title":"Elucidating gamma-ray induced effects at the copper-water interface for permanent disposal of spent nuclear fuels","authors":"Nakkyu Chae ,&nbsp;Inna L. Soroka ,&nbsp;Richard I. Foster ,&nbsp;Dokyu Kang ,&nbsp;Mats Jonsson ,&nbsp;Sungyeol Choi","doi":"10.1016/j.jnucmat.2025.155993","DOIUrl":"10.1016/j.jnucmat.2025.155993","url":null,"abstract":"<div><div>Gamma radiation can have an influence on the integrity of the copper canisters used in deep geological repositories for isolating radioactive waste. Understanding the interactions between aqueous radiolysis products and container materials, particularly at the copper-water interface, is essential for assessing the canister integrity. This study investigates the gamma-radiation-induced products on copper specimens and water through experimental methods. Cu specimens were exposed to gamma radiation, and corrosion products were analysed using cathodic reduction, XPS, ICP-MS, and FT-IR. Results show that Cu₂O is the dominant corrosion product formed during irradiation. Pre-oxidized Cu specimens, especially those formed at evaluated temperatures (140 °C), exhibited less corrosion depths and much more homogeneous coloration on the surfaces compared to literature data of irradiated bare Cu specimens and pre-oxidized 90 °C Cu specimens, suggesting the possibility that high temperature pre-oxidation enhances corrosion resistance under irradiation conditions. Additionally, the enhanced formation of alkane species, such as CH₄, was observed in irradiated water, likely originated from the reduction of CO₂ and HCO₃⁻ by radiation-induced reducing agents (H, H₂, and e_aq⁻). This observation raises new questions about the chemical transformations occurring under irradiation. The findings highlight the importance of understanding initial Cu oxide layer properties and suggest that optimizing temperature and environmental conditions in DGRs can improve the long-term performance of Cu canisters. Future studies are encouraged to explore localized corrosion mechanisms in Cl⁻-rich environments and to further investigate the implications of alkane production on the chemical stability of DGR systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155993"},"PeriodicalIF":2.8,"publicationDate":"2025-06-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144338825","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":"Achieving ultrahigh strength and high thermal stability of Cu-15 W composite by hierarchical structure with bimodal particle size distribution","authors":"Hongyi Song ,&nbsp;Qingqing Liu ,&nbsp;Ningyu Li ,&nbsp;Daohan Yu ,&nbsp;Yuxiang Chen ,&nbsp;Yueming Huang ,&nbsp;Mingyang Li ,&nbsp;Yongqin Chang","doi":"10.1016/j.jnucmat.2025.155992","DOIUrl":"10.1016/j.jnucmat.2025.155992","url":null,"abstract":"<div><div>Cu-15 wt. % W composites were successfully produced via a feasible and simple strategy, achieving an excellent combination of ultrahigh strength and high thermal stability. These composites demonstrate potential as candidates for heat sink materials in the fusion reactor. In this work, Cu-15 wt. % W composites with bimodal particle size distribution were fabricated by conventional mechanical alloying (MA) and spark plasma sintering (SPS) followed by hot rolling and annealing processes. The ultimate tensile strength, total elongation, electrical conductivity and thermal conductivity are 709 MPa, 13.7 %, 60.8 %IACS and 291.3 W/(m·K), respectively. Cu-15 wt. % W composite also has excellent thermal stability with the ultimate tensile strength decreasing only 3.4 % even when the annealing temperature is increasing to 700 °C. The excellent performance of the composite is due to the fine, high-density and bimodal size distribution reinforced particles (fine W particles with the average size of 27.5 nm and ultrafine W particles with the average size of 4.6 nm) in Cu matrix. The large-sized particles stabilize the microstructure against the effect of elevated temperature by pinning grain boundaries. Forming small-sized particles offers higher density strengthening phases improving the storage capability of dislocations. The combination of bimodal particle size distribution acts as the very powerful inhibition on the movement of dislocations and grain boundaries. Our study provides an effective route to achieve high performance of Cu matrix composites.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155992"},"PeriodicalIF":2.8,"publicationDate":"2025-06-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144472335","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 Data-driven machine learning model for radiation-induced DBTT shifts in RAFM steels","authors":"Pengxin Wang ,&nbsp;G.M.A.M. El-Fallah","doi":"10.1016/j.jnucmat.2025.155984","DOIUrl":"10.1016/j.jnucmat.2025.155984","url":null,"abstract":"<div><div>This study develops a stacking ensemble machine learning model to predict ductile-to-brittle transition temperature (DBTT) in irradiated Reduced-Activation Ferritic-Martensitic (RAFM) steels. Using a dataset of 490 irradiation cases, the model integrates XGBoost, Random Forest (RF), Gradient Boosting Decision Tree (GBDT), and Multi-Layer Perceptron (MLP), achieving a high predictive accuracy (R² = 0.96) and outperforming individual models. The results highlight the significant influence of irradiation dose and temperature on DBTT. Beyond 30 dpa, defect accumulation causes a sharp DBTT increase, while irradiation temperature exhibits a nonlinear effect, peaking at 150–300 °C due to radiation-enhanced precipitation and declining above 350 °C as defect recovery improves ductility. Additionally, alloying elements play a crucial role: Ta and W help mitigate embrittlement, moderate Cr (4–6 wt. %) increases DBTT, and higher Cr levels (&gt;6 wt. %) reduce it at elevated temperatures. SHAP analysis reveals that W is particularly effective in reducing embrittlement in alloys with moderate Cr (6–9 wt. %) and low Ta, while higher Cr concentrations (&gt;6 wt. %) help stabilise DBTT at elevated temperatures. To enable practical alloy design, a genetic algorithm was combined with the model to optimise steel compositions under defined irradiation conditions (200–350 °C, 10 dpa). The approach successfully identified candidate alloys with predicted DBTT values below 50 °C. This study provides a robust predictive framework for understanding and optimising DBTT in irradiated RAFM steels, offering valuable insights into their performance in next-generation nuclear reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155984"},"PeriodicalIF":2.8,"publicationDate":"2025-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144481675","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":"Computational exploration and experimental validation of reduced- multi-principal element alloys with stable BCC phases and superior mechanical properties","authors":"Yuji Harada ,&nbsp;Shi Woo Lee ,&nbsp;Katsuaki Tanabe ,&nbsp;Hyoung Seop Kim","doi":"10.1016/j.jnucmat.2025.155989","DOIUrl":"10.1016/j.jnucmat.2025.155989","url":null,"abstract":"<div><div>Structural materials for fusion reactors must possess exceptional mechanical strength to withstand extreme high-temperature and irradiation conditions. Multi-principal element alloys (MPEAs) have emerged as promising candidates due to their exceptional irradiation resistance and high-temperature mechanical properties. The reduced-activation type MPEAs (RAMPEAs) are particularly noted for mitigating the effects of radioactivation and contributing to the safe operation of fusion reactors and waste management. This study systematically explored the compositional regimes of ternary, quaternary, and quinary RAMPEAs composed of nine low-activation elements. Using a yield strength prediction model, we performed phase diagram calculations and employed the valence electron concentration as a ductility index, screening a total of 625,518 compositions. As a result, 83 promising chemical compositions were identified on the Pareto front based on three objective functions: yield strength at 1000 K, valence electron concentration, and phase stability. Among them, Ti₄₀V₂₀Cr₃₀W₁₀, Ti₄₀V₁₅Fe₁₀Zr₃₅, and Ti₄₀Fe₂₅Zr₃₅ were experimentally evaluated for their predicted superior balance of yield strength, ductility, and phase stability. The computational framework was validated through compression tests and microstructure analysis, achieving a yield strength of 1.39 GPa and a fracture elongation of 16.5 % for Ti₄₀V₂₀Cr₃₀W₁₀ at room temperature, surpassing most RAMPEAs reported in previous studies. This study provides valuable insights into the RAMPEA design and highlights promising chemical compositions for nuclear fusion reactor applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155989"},"PeriodicalIF":2.8,"publicationDate":"2025-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144480750","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 Fe2+ ion-irradiation on additively manufactured Inconel 617 alloy","authors":"Tomasz Stasiak ,&nbsp;Jarosław Jasiński ,&nbsp;Józef Rzempołuch ,&nbsp;Udisien Woy ,&nbsp;Magdalena Wilczopolska ,&nbsp;Katarzyna Mulewska ,&nbsp;Marcin Kowal ,&nbsp;Katarzyna Ciporska ,&nbsp;Łukasz Kurpaska ,&nbsp;Jacek Jagielski","doi":"10.1016/j.jnucmat.2025.155978","DOIUrl":"10.1016/j.jnucmat.2025.155978","url":null,"abstract":"<div><div>Inconel 617 alloy is a leading candidate for the intermediate heat exchanger in Generation IV nuclear reactors. Although Inconel 617 is well known and applied in many engineering fields, the irradiation resistance and preparation by additive manufacturing are rarely studied. The aim of this paper is to investigate the ion irradiation resistance of additively manufactured Inconel 617. The alloys were prepared either from wire or powder feedstocks using a directed energy deposition technique. Then, the samples were submitted to different post-processing protocols. The irradiation was carried out using Fe²⁺ ions at room temperature with an energy of 10 MeV up to a dose of approximately 1 dpa. The microstructure of the samples was investigated by SEM-EDS and TEM, while the irradiation effects were evaluated by TEM and nanoindentation. The microstructural investigations reveal some differences in the microstructure of the studied samples. The samples produced from powder feedstock show larger grains and a higher volume fraction of precipitates than those from wire feedstock. The annealing followed by water quenching decreases the volume fraction of precipitates, while hot isostatic pressing induces the elongation of grains. Despite microstructural differences, the hardness of all studied samples is quite similar. The ion-irradiation experiments show slight differences in the behavior of the analyzed samples. The nanoindentation experiments reveal irradiation-induced hardening by 21.9–36.8 %, depending on the sample. TEM instigations show the presence of defect clusters induced by ion irradiation in the peak damage area.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155978"},"PeriodicalIF":2.8,"publicationDate":"2025-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144297958","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":"Properties and applications of deuterium absorption-desorption in TiZrHfNbTa high entropy alloy","authors":"Zhaorui Chen ,&nbsp;Siwei Zhang ,&nbsp;Xingyu Jiang ,&nbsp;Mingjie Zheng ,&nbsp;Hao Yang ,&nbsp;Yongju Sun ,&nbsp;Gang Li","doi":"10.1016/j.jnucmat.2025.155987","DOIUrl":"10.1016/j.jnucmat.2025.155987","url":null,"abstract":"<div><div>High-entropy alloys (HEAs) demonstrate exceptional potential in nuclear applications due to their comprehensive advantages in radiation resistance, hydrogen (isotope) storage capacity, hydrogen (isotope) absorption kinetics, and structural stability. Addressing the issues of traditional neutron target materials, such as poor hydrogen absorption kinetics, low hydrogen desorption temperatures, and low storage capacity, this study melted a body-centered cubic (BCC) TiZrHfNbTa HEA. The deuterium absorption-desorption characteristics and microstructural evolution of the TiZrHfNbTa HEA were investigated, demonstrating its potential as a material for neutron target applications. Compared to titanium, TiZrHfNbTa exhibited different kinetic properties and slightly higher deuterium storage capacity, with a deuterium-to-metal ratio exceeding 1.9 and higher desorption temperatures. The alloy maintained stable deuterium capacity after ten cycles, showing good cyclic stability. This work extends the application of high-entropy alloys in the nuclear energy field.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155987"},"PeriodicalIF":2.8,"publicationDate":"2025-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502059","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":"Simulation of tritium release processes from lithium ceramics under reactor irradiation conditions with varying reactor power","authors":"Timur Kulsartov ,&nbsp;Inesh Kenzhina ,&nbsp;Yergazy Kenzhin ,&nbsp;Yevgeniy Chikhray ,&nbsp;Asset Shaimerdenov ,&nbsp;Zhanna Zaurbekova ,&nbsp;Alexander Yelishenkov ,&nbsp;Kuanysh Samarkhanov ,&nbsp;Meiram Begentayev ,&nbsp;Saulet Askerbekov ,&nbsp;Aktolkyn Tolenova","doi":"10.1016/j.jnucmat.2025.155979","DOIUrl":"10.1016/j.jnucmat.2025.155979","url":null,"abstract":"<div><div>It is known that one of the most important parameters for lithium ceramics, promising for use in solid-state fusion blankets, includes their properties regarding tritium. Tritium generated in lithium ceramics (as a result of reaction of ⁶Li atom with neutron) is released from the surface after passing through a chain of processes:</div><div>1. it is initially thermalized in the ceramic volume (it should be noted that at this stage, if the reaction of its generation has occurred in the near-surface region, it can be released from the sample in an inactivation-free manner),</div><div>2. after that it diffuses to the open surface, interacting with traps in the ceramic volume (here both reversible and non-reversible capture of tritium by traps, which are usually defects in the structure, is possible);</div><div>3. tritium atoms associate with each other or with hydrogen impurities on the surface, after which the resulting molecules are desorbed.</div><div>To determine the parameters of gas release from ceramics generated by neutron irradiation of tritium, reactor studies are usually carried out. Analysis of the results of such reactor experiments is usually complicated by the need to consider various factors and requires careful consideration of these factors while creating and selecting model parameters.</div><div>The present work presents data on modeling the results of a previously conducted reactor experiment. Modeling was carried out by the finite element method (FEM) based on a complex model that considers temperature gradients across the ceramics during the reactor experiment, as well as the processes of tritium diffusion in the volume and desorption of tritium molecules from the ceramic surface. We investigated the sections of the reactor experiment where the reactor power was varied, leading to changes in the temperature fields in the ceramic samples and the released flux of tritium molecules.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155979"},"PeriodicalIF":2.8,"publicationDate":"2025-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144306230","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":"Experimental characterizations by EBSD and TEM of sub-grain boundaries and dislocations in low irradiated UO2 fuels","authors":"C. Onofri,&nbsp;I. Zacharie-Aubrun,&nbsp;D. Drouan,&nbsp;T. Blay,&nbsp;C. Sabathier,&nbsp;T. Grenèche,&nbsp;K. Hanifi,&nbsp;C. Rodriguez,&nbsp;L. Fayette,&nbsp;S. Reboul,&nbsp;N. Robert,&nbsp;P. Binet,&nbsp;M. Bertolus","doi":"10.1016/j.jnucmat.2025.155981","DOIUrl":"10.1016/j.jnucmat.2025.155981","url":null,"abstract":"<div><div>Transmission electron microscopy (TEM) and electron backscattered diffraction (EBSD) characterizations of irradiated UO₂ fuels at low burn-ups (sample burn-ups of 16 and 37 GWd/t_U) are conducted. Weakly disoriented sub-grain boundaries (distribution and misorientations) are studied by EBSD at several radial positions for both samples. They are observed close to initial grain boundaries all along the pellet radius, since one cycle of irradiation. They present very low misorientations. Average misorientations in grains increase with burn-up. As observed at higher burn-ups, a small decrease at mid-radius in misorientations is also evidenced for each sample. The microstructure at 0.56R is characterized in detail by TEM for both samples. Isolated dislocations (loops and lines) are observed within the grains as well as dislocation networks (forming low disoriented sub-grain boundaries) located near initial grain boundaries. Dislocation loops and line densities, motion mechanisms and networks characteristics are obtained from TEM image analysis. The results highlight the complexity of dislocation motion in the irradiated fuels, mainly through mixed climb mechanisms at 16 GWd/t_U, and by mixed climb and slip (mainly in the easiest system of UO₂: ½&lt;110&gt;{100}) at 37 GWd/t_U. Some discrepancies with the literature concerning dislocation density changes with burn-up are also highlighted. The line density seems to be constant whereas the loop density tends to increase. At 0.56R, sub-grain boundaries correspond mainly to symmetrical tilt sub-grain boundaries or networks of entangled dislocation lines.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155981"},"PeriodicalIF":2.8,"publicationDate":"2025-06-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144330617","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":"Production of anhydrous f-element fluorides through the ionothermal treatment of f-element oxalates","authors":"Elodia Ciprian ,&nbsp;Bryan J. Foley ,&nbsp;Tyler L. Spano ,&nbsp;Andrew Miskowiec ,&nbsp;Thomas Shehee ,&nbsp;Amy E. Hixon","doi":"10.1016/j.jnucmat.2025.155983","DOIUrl":"10.1016/j.jnucmat.2025.155983","url":null,"abstract":"<div><div>The pivotal role of uranium and plutonium fluorides in the nuclear fuel cycle, particularly in the pyrochemical reduction process, is well recognized. Traditionally, the fluorination of uranium and plutonium materials relies on the use of highly toxic and corrosive gases (e.g., HF_(g), F_2(g)). Herein, we present an alternative approach using the ionic liquid 1‑butyl‑3-methylimidazolium hexafluorophosphate ([Bmim][PF₆]_(l)) and/or hexafluorophosphoric acid (HPF_6(aq)) as fluorinating agents for the <em>f</em>-element oxalates <span><math><mrow><msubsup><mi>M</mi><mn>2</mn><mtext>III</mtext></msubsup><msub><mrow><mo>(</mo><mrow><msub><mi>C</mi><mn>2</mn></msub><msub><mi>O</mi><mn>4</mn></msub></mrow><mo>)</mo></mrow><mn>3</mn></msub><mo>·</mo><mn>9</mn><msub><mi>H</mi><mn>2</mn></msub><mi>O</mi><mrow><mo>(</mo><mi>s</mi><mo>)</mo></mrow><msubsup><mi>M</mi><mn>2</mn><mtext>III</mtext></msubsup><msub><mrow><mo>(</mo><mrow><msub><mi>C</mi><mn>2</mn></msub><msub><mi>O</mi><mn>4</mn></msub></mrow><mo>)</mo></mrow><mn>3</mn></msub><mo>·</mo><mn>9</mn><msub><mi>H</mi><mn>2</mn></msub><msub><mi>O</mi><mrow><mo>(</mo><mi>s</mi><mo>)</mo></mrow></msub></mrow></math></span> (M^III = Ce, Pu) and <span><math><mrow><msup><mrow><mi>M</mi></mrow><mtext>IV</mtext></msup><msub><mrow><mo>(</mo><mrow><msub><mi>C</mi><mn>2</mn></msub><msub><mi>O</mi><mn>4</mn></msub></mrow><mo>)</mo></mrow><mn>2</mn></msub><mo>·</mo><mn>6</mn><msub><mi>H</mi><mn>2</mn></msub><msub><mi>O</mi><mrow><mo>(</mo><mi>s</mi><mo>)</mo></mrow></msub></mrow></math></span> (M^IV = Th, U). Our findings demonstrate that [Bmim][PF₆]_(l) and HPF_6(aq) enable the ionothermal fluorination of <em>f</em>-element oxalates, resulting in the formation of anhydrous CeF_3(s), ThF_4(s), and UF_4(s) within 2 hours at 200 °C. This method also facilitates the partial fluorination of plutonium(III) oxalate, yielding a mixture of anhydrous PuF_3(s) and an unidentified phase. Overall, the ionothermal treatment approach offers a safer and more efficient means of producing anhydrous <em>f</em>-element fluorides than conventional methods involving hazardous gases. In addition, we describe the morphology of UF_4(s) materials as a function of production route and demonstrate the presence of morphological signatures that could be used during a nuclear forensic investigation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155983"},"PeriodicalIF":2.8,"publicationDate":"2025-06-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144330618","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":"Mechanism of neutron irradiation-induced surface crack density changes in Chromium-coated zirconium alloys","authors":"Mingda Han ,&nbsp;Weixu Zhang ,&nbsp;Junnan Lv ,&nbsp;Hongyan Yang ,&nbsp;Jishen Jiang","doi":"10.1016/j.jnucmat.2025.155982","DOIUrl":"10.1016/j.jnucmat.2025.155982","url":null,"abstract":"<div><div>This study investigates the effect of irradiation on the crack density of chromium-coated zirconium alloys. Based on the continuum mechanics framework, irradiation damage is described through changes in the stress-strain curve. A fracture prediction method for chromium coatings is established using a strength theory and the birth and death element method, showing good agreement with experimental trends. The study shows that with increasing irradiation dose, the substrate material undergoes irradiation-induced embrittlement, leading to a significant increase in interfacial shear stress. As a result, the initial cracking strain decreases from approximately 0.414 % to 0.408 %, the saturated crack density increases by about 36.4 %, and the crack opening displacement decreases by around 36.6 %. For coating thickness, a thicker coating leads to a lower saturated crack density. When the substrate thickness exceeds the critical value, it no longer affects the saturated crack density. The residual stress in the coating does not affect the saturated crack density but only influences the crack initiation strain. This study provides theoretical support for the irradiation-resistant design of accident-tolerant fuel cladding coatings.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155982"},"PeriodicalIF":2.8,"publicationDate":"2025-06-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144313585","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}