Journal of Nuclear Materials最新文献

{"title":"Study on long-term alteration behavior of simulated sulfate-bearing HLLW waste glass under thermal–hydrological–mechanical–chemical multi-field conditions","authors":"Zhengdi Jiang ,&nbsp;Xiaolin Yin ,&nbsp;Jiaxin Huang ,&nbsp;Ya Li ,&nbsp;Liguo Xu ,&nbsp;Lang Wu","doi":"10.1016/j.jnucmat.2025.156379","DOIUrl":"10.1016/j.jnucmat.2025.156379","url":null,"abstract":"<div><div>Glass waste forms are at risk of groundwater intrusion during long-term geological disposal, where direct contact compromises chemical durability and may release radionuclides into the biosphere, thus necessitating a critical assessment of their chemical stability in aqueous environments. This study investigated the chemical stability of simulated sulfate-bearing high-level liquid waste (HLLW) glass under thermal–hydrological–mechanical–chemical (THMC) multi-field conditions (90°C, 0.01 mL/min flow rate, 10 MPa, in simulated groundwater) through 364-day multi-stage leaching tests. Results revealed sequential precipitation of platy BaSO₄ (7–14 days), Mg-Al-rich layered silicate (at 14 days), and acicular/prismatic CaCO₃ crystals (by 364 days). Alteration layer development initiated between 14 and 56 days (reaching 23 μm by 56 days) and thickened to 135.6 μm by 364 days, comprising three distinct zones: an innermost amorphous aluminosilicate gel layer, Mg-Al-rich silicates (containing BaSO₄), and an outermost CaCO₃ layer observed at 364 days. Dissolution rates exhibited a multi-stage evolution: rapid increase (1–3 days), decelerated increase (3–14 days), sharp decline (14–56 days), a stabilization trend (56–182 days), and the near-achievement of dissolution equilibrium (182–364 days). These findings offer important insights into the evolution of waste glass alteration under THMC multi-field conditions, yielding key safety assessment data for high-level radioactive waste disposal.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156379"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789293","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":"Immobilization of high-sodium and cesium-rich waste derived from TRPO process in single phase hollandite ceramic waste forms","authors":"Zhiguo Yang ,&nbsp;Guoqiang Zhao ,&nbsp;Wanjun Shi ,&nbsp;Xianzhou Ning ,&nbsp;Bo Xie ,&nbsp;Wei Zhang ,&nbsp;Bin Ye ,&nbsp;Yushan Yang","doi":"10.1016/j.jnucmat.2025.156368","DOIUrl":"10.1016/j.jnucmat.2025.156368","url":null,"abstract":"<div><div>A single phase hollandite waste form was developed to immobilize the high-sodium and cesium-rich waste (HSCRW) stream derived from the trialkyl phosphine oxide (TRPO) process. In this work, the (1-<em>x</em>)Ba_1.2Cr_2.4Ti_5.6O₁₆·<em>x</em>HSCRW (0.0 ≤ <em>x</em> ≤ 0.2) ceramics were fabricated to investigate the effect of HSCRW incorporation on phase composition, microstructure and chemical durability of the synthesized hollandite ceramics. It was found that all waste elements are successfully embedded into the hollandite crystal structure, and the samples sintered at 1150 °C with <em>x</em> ≤ 0.15 showed a pure hollandite phase. The leaching test indicated that the normalized leaching rates of the waste elements Cs, Na, Rb, Sr, Mo, Fe, Ni, Ru and Rh in the as-prepared ceramic waste forms were ∼ 10^-3 <em>g</em>·m^-2·d^-1, with corresponding <em>LX</em> values &gt; 14.5 after 28 days of leaching. Moreover, the leached samples maintained a single-phase hollandite with tetragonal structure (<em>I</em>4/m). These results demonstrate that hollandite ceramics can serve as promising host matrices for immobilizing HSCRW.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156368"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789374","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":"Evaluation of damage effects in 304 high-boron austenitic stainless steel under only and simultaneous Fe, He ions irradiation","authors":"Jinhua Hao ,&nbsp;Peng Zhang ,&nbsp;Xi Li ,&nbsp;Bin Wang ,&nbsp;Zhihui Cai ,&nbsp;Lifeng Ma ,&nbsp;Xianxiu Mei","doi":"10.1016/j.jnucmat.2025.156378","DOIUrl":"10.1016/j.jnucmat.2025.156378","url":null,"abstract":"<div><div>The introduction of precipitates into austenitic stainless steel is an effective strategy for improving irradiation resistance. This study employed only and simultaneous Fe, He ions irradiation on 304 high-boron austenitic stainless steel at room temperature and investigated the mechanism of ion irradiation on irradiation-induced defects and mechanical property. Throughout the ion penetration region under different irradiation conditions, both the austenite and TiB₂ phase retained original crystal structure, while the (Fe,Cr)₂B phase underwent a crystal-to-amorphous transformation. Only He ions irradiation generated nanoscale helium bubbles and induced a high density of fine dislocation loops in the damage peak region within the austenitic matrix. Under simultaneous Fe+He ions irradiation, the average size of helium bubbles and the width of helium bubbles layer increased, and at the same time, the incorporation of Fe ions also promoted the coalescence and growth of dislocation loops. The amorphization of the (Fe,Cr)₂B phase after ion irradiation played a dominant role in softening, while the pinning effect of helium bubbles inhibited hardness decrease. Irradiation-induced dislocation loops played a major role in hardening, especially the small and dense dislocation loops induced by only He ions irradiation generated significant hardening effects. Under simultaneous Fe+He ions irradiation, the 304 high-boron stainless steel with diffusely distributed borides exhibited slight softening without evident hardening. The behavior demonstrated superior resistance to irradiation hardening compared to conventional austenitic stainless steel.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156378"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789383","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":"Combined effect of Te and impurities in molten LiF-NaF-KF salt on the corrosion of 316H SS and UNS N10003 alloy","authors":"Miaomiao Hu ,&nbsp;Xinmei Yang ,&nbsp;Huajian Liu ,&nbsp;Xingtai Zhou","doi":"10.1016/j.jnucmat.2025.156397","DOIUrl":"10.1016/j.jnucmat.2025.156397","url":null,"abstract":"<div><div>The corrosion mechanisms of nickel-based alloy (UNS N10003) and stainless steel (316H SS) in molten LiF-NaF-KF (FLiNaK) salt at 700 °C were investigated. Results indicate that UNS N10003 exhibits better resistance to intergranular corrosion than 316H SS in the molten FLiNaK salt with and without Te. In the same batch of molten FLiNaK salt without Te, for UNS N10003 alloy, the reactions between its alloying elements and the impurities in the salt caused the uniform corrosion; while the reactions led to an intergranular corrosion for 316H SS. The presence of Te in molten FLiNaK salt induced the intergranular corrosion for UNS N10003 alloy. The coexistence of Te (1 wt% Te) and the impurities in the molten FLiNaK salt aggravated the corrosion of UNS N10003 alloy and 316H SS by ∼12 times. The corrosion depth of 316H SS (∼370 μm for 400 h) is larger than that of UNS N10003 (∼90 μm for 400 h) in the molten FLiNaK salt with Te (1 wt%). The severer intergranular corrosion is mainly attributed to the reactions involving Te, the impurities in salt, the alloying elements, and the precipitates at grain boundaries.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156397"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145838099","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":"Leaching behavior of HLW glass waste form in Beishan groundwater environment","authors":"Feng Zhiqiang ,&nbsp;Wang Ju ,&nbsp;Xie Jingli ,&nbsp;Cheng Jianfeng ,&nbsp;Lin Jie ,&nbsp;Xie Hua","doi":"10.1016/j.jnucmat.2025.156373","DOIUrl":"10.1016/j.jnucmat.2025.156373","url":null,"abstract":"<div><div>In 2021, China's first high-level radioactive waste vitrification facility commenced operation in Guangyuan, Sichuan, while the Beishan Underground Laboratory in Gansu initiated construction. This study investigates the glass waste forms produced domestically, analyzing their surface characteristics and elemental release tendencies in both deionized water and the complex hydrogeochemical environment of Beishan groundwater, which is crucial for assessing long-term disposal safety. Research results demonstrate that glass corrosion mechanisms differ significantly between deionized water and complex Beishan groundwater. In deionized water, corrosion proceeds primarily via relatively simple ion exchange and network hydrolysis. In contrast, the complex ionic environment of Beishan groundwater triggers active interface reactions, leading to the formation of various silicate precipitation layers. These layers introduce a surface \"blocking-and-release\" effect, making the apparent leaching behavior more complex. The initial leaching path depends on the glass's surface condition. Alkali metals enriched on as-cast samples promote simple MgO phase formation, while the rougher surface of processed samples facilitates rapid growth of complex silicates. Despite different initial paths, surface layer evolution converges after long-term leaching. Leaching rates for matrix elements rapidly decreased to 10^–1 g·m^-2·d^-1 by day 14, then slowed to 10^–2 g·m^-2·d^-1 by day 92. The trivalent simulant La exhibited a much lower and faster-declining rate, dropping to 10^–3 g·m^-2·d^-1 by day 7 and remaining at that level thereafter, showing excellent immobilization effect on actinide elements.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156373"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789292","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":"Performance of W and W-Re alloys under low-energy He+ ion irradiations at relevant nuclear fusion environments","authors":"Tyler E. Ray,&nbsp;Jitendra K. Tripathi,&nbsp;Jack C. Johnson,&nbsp;Ahmed Hassanein","doi":"10.1016/j.jnucmat.2025.156278","DOIUrl":"10.1016/j.jnucmat.2025.156278","url":null,"abstract":"<div><div>Fusion-relevant tungsten (W) and tungsten-rhenium (W-Re) alloys were exposed to low-energy helium (He) ions at divertor base temperatures relevant to ITER/DEMO conditions. Pure W and W-Re alloys (1, 3, 5 wt. % Re) were irradiated with 100 eV He⁺ ions (fluence ∼<span><math><mrow><mn>1.08</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mn>24</mn></msup></mrow></math></span> ions m⁻²) from 900 to 1300 K. Extensive analysis utilized scanning electron microscopes (SEM) with focused ion beam (FIB) cross-sectioning, atomic force microscope (AFM), optical reflectivity, x-ray diffraction (XRD), and x-ray photoelectron spectroscopy (XPS) to investigate the surface evolution. Early-stage fuzz formed on all samples at 900 K, with bubble size increasing with Re content. At 1300 K, W-3Re and W-5Re exhibited surface roughening and pore formation, but no evidence of fuzz growth. Reflectivity increased significantly at 1300 K, with W-3Re reaching ∼47 %, higher than W-5Re, indicating less surface roughening with 3 wt. % Re. FIB cross-sectioning at 1100 K revealed a fuzz depth of ∼600 nm on both pure W and W-5Re samples, where under these experimental conditions of intermediate temperature, fuzz growth does not depend on Re concentration. X-ray diffractograms showed consistent crystallographic orientations for pure W and W-5Re. Microstrain analysis (Williamson-Hall) method revealed an increase of ∼0.070 ± 0.002 % in the annealed surface of W-5Re compared to pure W. Post-irradiation the microstrain increased significantly, with pure W exhibiting approximately threefold measured increase in microstrain compared to W-5Re, highlighting the stabilizing effect of substitutional Re. XPS results confirmed the Re content in the alloy and revealed an increased oxide percentage on more developed fuzzy surfaces. These results suggest potential mechanisms for W-Re alloys observed decrease in fuzz forming temperature window, providing crucial insights into future materials to potentially extend the divertor lifetime.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"619 ","pages":"Article 156278"},"PeriodicalIF":3.2,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145517591","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":"Interstitial-mediated precipitation unveiled: Mechanism of irradiation-induced rhenium aggregation in molybdenum","authors":"Xiang-Qian Wu ,&nbsp;Wei Zhao ,&nbsp;Yu-Hao Li ,&nbsp;Zhi-Peng Sun ,&nbsp;Yong Xin ,&nbsp;Xi Qiu ,&nbsp;Huiqiu Deng ,&nbsp;Hong-Bo Zhou ,&nbsp;Guang-Hong Lu","doi":"10.1016/j.jnucmat.2025.156283","DOIUrl":"10.1016/j.jnucmat.2025.156283","url":null,"abstract":"<div><div>We investigated the mechanisms of rhenium (<em>Re</em>) aggregation and precipitation in molybdenum (Mo) under irradiation by employing a combination of first-principles calculations and object kinetic Monte Carlo (OKMC) simulations. Our results demonstrate that irradiation-induced defects significantly affect the distribution of <em>Re</em>, resulting in distinct patterns compared to non-irradiated conditions. In pristine Mo, <em>Re</em> atoms favor the even distribution due to the repulsive interactions. In contrast, strong attractive interactions are observed between <em>Re</em> and defects, particularly self-interstitial atoms (SIAs). Furthermore, the rapid three-dimensional diffusion of <em>Re</em>-SIA complexes in Mo facilitates their efficient trapping by substitutional <em>Re</em> atoms, leading to the formation of stable <em>Re</em>-<em>Re</em> dumbbells. These dumbbells subsequently serve as effective traps for subsequent <em>Re</em> atoms, thereby driving the nucleation and growth of <em>Re</em>-rich clusters. Based on these findings, we propose an interstitial-mediated migration and aggregation mechanism for the irradiation-induced <em>Re</em> precipitation in Mo. Building on these insights, we utilized OKMC simulations to examine the co-evolution of <em>Re</em> and irradiation-induced defects in Mo. The simulations reveal that <em>Re</em> atoms aggregate to form <em>Re</em>-rich clusters, and eventually precipitates, with increasing irradiation dose. Moreover, the number of <em>Re</em> atoms in interstitial-type clusters surpasses that in vacancy-type clusters, corroborating the interstitial-mediated mechanism. These results provide a valuable reference for the composition optimization and property evaluation of Mo-<em>Re</em> alloys under irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"619 ","pages":"Article 156283"},"PeriodicalIF":3.2,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145517590","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 insights into dislocation–void interactions in tungsten–vanadium alloys under fusion-Relevant conditions","authors":"Ayobami D. Daramola","doi":"10.1016/j.jnucmat.2025.156319","DOIUrl":"10.1016/j.jnucmat.2025.156319","url":null,"abstract":"<div><div>Tungsten–vanadium (W–V) alloys are promising candidates for fusion plasma-facing components, yet irradiation-induced nanoscale voids can strongly hinder dislocation glide and degrade mechanical performance. Here we use large-scale molecular-dynamics simulations to examine edge-dislocation motion and void bypass in single-crystal W and random W–V alloys containing 5 and 10 at.% V over a fusion-relevant temperature range (300–900 K) and void diameters of 2–5 nm. Vanadium reduces the unstable stacking-fault energy anisotropically while preserving high-temperature stiffness, and introduces substantial solute drag that elevates the CRSS for obstacle-free glide, particularly at intermediate temperatures. In the presence of voids, bypass occurs through a reproducible sequence involving screw-segment nucleation, prismatic loop emission, Orowan bow-out, and jog-mediated release. The stresses associated with these stages increase with V content and diminish with temperature. The peak bypass stress, <em>τ</em>_max(<em>d</em>_eff, <em>T</em>), is described accurately by a modified Bacon–Kocks–Scattergood line-tension formulation in which geometric scaling is retained and solute and thermal effects enter as nearly additive offsets. The resulting obstacle strengths, <em>α</em>, for 2 nm voids compare favourably with recent in-situ TEM measurements at 293 and 723 K. Collectively, these results provide stage-resolved mechanistic insight, deliver continuum-ready hardening parameters, and identify  ∼ 10 at.% V as a composition that offers a favourable balance between enhanced void resistance and preserved post-bypass mobility in tungsten-based fusion materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156319"},"PeriodicalIF":3.2,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145681994","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 the incorporation capacity of calcine of simulated high-level liquid waste in borosilicate glass matrix","authors":"Wang Changfu ,&nbsp;Liu Xu ,&nbsp;Li Lin ,&nbsp;Li Wengzheng ,&nbsp;Shao Bobo ,&nbsp;Liu Yan ,&nbsp;Wang Yun ,&nbsp;Huang Xi ,&nbsp;Tan Shengheng ,&nbsp;Liu Zhirong ,&nbsp;Zhang Shengdong","doi":"10.1016/j.jnucmat.2025.156208","DOIUrl":"10.1016/j.jnucmat.2025.156208","url":null,"abstract":"<div><div>As a pivotal technology for high-level liquid waste (HLW) treatment, vitrification plays a crucial role in ensuring the sustainable development of nuclear energy. The two-step vitrification process, which involves initial calcination of HLW into solid calcine followed by melting with base glass, has demonstrated significant improvements in immobilization efficiency. This study investigated the capacity of simulated HLW calcine to be incorporated within a borosilicate glass matrix.Through comprehensive characterization techniques including field emission scanning electron microscopy coupled with energy dispersive spectroscopy (SEM-EDS), X-ray diffraction (XRD), Fourier-transform infrared spectroscopy (FTIR), and Raman spectroscopy, the effects of calcine content (18–32 wt%) on the microstructure, chemical composition, phase distribution, and network structure of the vitrified matrixs were examined. Homogeneous glass matrices are achieved at 18–26 wt% calcine content, while phase separation and crystalline precipitation occur at higher concentrations. However, when the content reaches 28 wt%, the initial precipitation of granular CaMoO₄ crystals occurs, maintaining relatively uniform distribution. Further increasing the content to 30 wt% leads to the formation of acicular CaMoO₄ phases accompanied by minor Zr_xCe_1-xO₂ particle precipitation. Notably, at 32 wt% calcined product content, a distinct phase separation layer emerges on the surface of the vitrified matrix, with a substantial increase in Zr_xCe_1-xO₂ precipitates. These research findings have clarified the incorporation capacity limits of simulated HLW calcine in borosilicate glass matrix, providing valuable theoretical guidance for enhancing waste loading in vitrified forms during the vitrification process.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156208"},"PeriodicalIF":3.2,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321510","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":"Revealing the superior SCC resistance of a dual-phase stainless steel in simulated PWR primary water via high-resolution characterization","authors":"Jun Xiao ,&nbsp;Yuhao Zhou ,&nbsp;Ting Xiao ,&nbsp;Cheng Peng ,&nbsp;Hao Wang ,&nbsp;Jiamei Wang ,&nbsp;Kai Chen","doi":"10.1016/j.jnucmat.2025.156317","DOIUrl":"10.1016/j.jnucmat.2025.156317","url":null,"abstract":"<div><div>The stress corrosion cracking (SCC) resistance of a Z3CN20–09 M duplex stainless steel (DSS) was systematically studied and compared to conventional single-phase austenitic stainless steel (316NG) under simulated pressurized water reactor (PWR) primary water conditions. Results show that the DSS exhibits much lower SCC growth rates than the 316NG in either deformed or high corrosion potential conditions. High-resolution microstructural characterization revealed the oxidation and crack-tip behaviors during SCC tests in simulated primary water, while in-situ SEM tensile experiments elucidated the deformation compatibility between austenite and ferrite phases, confirming that the duplex structure fundamentally enhances SCC resistance. The ferrite phase, which contains higher Cr, appears to promote the formation of protective chromium-rich oxide layers, thereby reducing oxidation-driven crack propagation. Additionally, ferrite shows lower strain accumulation and appears to limit crack propagation at austenite-ferrite interfaces.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156317"},"PeriodicalIF":3.2,"publicationDate":"2026-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145616126","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}