Journal of Nuclear Materials最新文献

{"title":"The effect of zirconium-microalloying on hydrogen diffusion behavior and hydride phase transformation in yttrium: A first principles study","authors":"Nan Zhou ,&nbsp;Chengqin Zou ,&nbsp;Guiyang Huang ,&nbsp;Baoqin Fu ,&nbsp;Xunxiang Hu","doi":"10.1016/j.jnucmat.2025.155919","DOIUrl":"10.1016/j.jnucmat.2025.155919","url":null,"abstract":"<div><div>Yttrium hydride has been perceived as an excellent neutron moderator for high-temperature nuclear reactor environments, owing to the high density of hydrogen (similar to neutron mass) and superior thermal stability. There is a growing demand for large-scale, crack-free bulk yttrium hydride for the micro-nuclear reactor applications. However, fabricating crack-free yttrium hydride with desired geometry and configuration is challenging and the relatively high neutron absorption cross section of yttrium prevents its widespread use. Zirconium microalloying offers a promising solution by mitigating cracking during hydriding and enhancing neutronic performance due to zirconium’s low neutron absorption cross section, thus broadening the application of yttrium-based materials. A fundamental understanding of atomic-scale interactions between hydrogen atoms and alloying elements is essential for optimizing the processing of yttrium-based alloys and mitigate hydrogen-induced cracking. In this study, the first-principles calculations were employed to investigate hydrogen diffusion behavior and hydrogen-induced phase transformation in pure yttrium and yttrium-zirconium alloys. The results show that an optimal concentration of zirconium enhances the diffusion coefficient of interstitial hydrogen in the bulk material. Moreover, the preferred phase transformation pathway for yttrium hydride was identified, as {0001}_HCP/{111}_FCC &amp; &lt;1<span><math><mover><mn>2</mn><mo>¯</mo></mover></math></span>10&gt;_HCP/&lt;<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>10&gt;_FCC. These findings highlight the pivotal role of hydrogen in phase transformation process and demonstrate that low zirconium concentrations promote the phase transformation of yttrium hydride. The mechanistic insights gained will aid in the development of yttrium-based alloys for high-temperature moderators in advanced nuclear reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155919"},"PeriodicalIF":2.8,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144146733","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":"Assessing shallow cracking in copper pipes under various conditions","authors":"H.M. Shirazi ,&nbsp;M. Golozar ,&nbsp;K. Daub ,&nbsp;J. Turnbull ,&nbsp;J.D. Giallanardo ,&nbsp;M. Behazin ,&nbsp;P.G. Keech ,&nbsp;M.R. Daymond ,&nbsp;S.Y. Persaud","doi":"10.1016/j.jnucmat.2025.155918","DOIUrl":"10.1016/j.jnucmat.2025.155918","url":null,"abstract":"<div><div>The corrosion in two ex-service copper pipes removed from the shutdown National Research Universal (NRU) reactor after 40 years of irradiation was investigated, focusing on localized corrosion. One pipe was exposed to humid air with a temperature of around 40 °C and a gamma radiation dose rate of 0.015 Gy/h, while the other sample was remote from the reactor and almost non-irradiated. Nanoscale analysis revealed shallow cracking and nitrogen at the crack tips of both the irradiated and non-irradiated cross-sections. However, the low-dose irradiation caused propagation of the pre-existing cracks observed in all Cu piping and did not initiate these defects. This research shows that stress corrosion cracking is unlikely to be enhanced in a defect-free electrodeposited copper under low-dose irradiation, similar to that expected for used fuel containers in Canada, in a deep geological repository.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155918"},"PeriodicalIF":2.8,"publicationDate":"2025-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144146732","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":"Increasing Tc retention during nuclear waste vitrification: Effect of alumina source on the rate of rhenium incorporation into glass","authors":"Laura Löwy ,&nbsp;Matouš Eret ,&nbsp;Jaroslav Kloužek ,&nbsp;Petra Cincibusová ,&nbsp;Miroslava Vernerová ,&nbsp;Martina Kohoutková ,&nbsp;Jaime George ,&nbsp;Melanie Killmer ,&nbsp;Pavel Ferkl ,&nbsp;Pavel Hrma ,&nbsp;Albert A. Kruger ,&nbsp;Richard Pokorný","doi":"10.1016/j.jnucmat.2025.155916","DOIUrl":"10.1016/j.jnucmat.2025.155916","url":null,"abstract":"<div><div>Technetium (Tc) volatilization during nuclear waste vitrification poses a significant challenge, necessitating recycle loops that may lead to reduced waste loading in the glass, or secondary Tc treatment in alternative waste forms. Thus, reducing the volatilization of Tc (or Re, its non-radioactive surrogate) is crucial for improving vitrification process efficiency and minimizing the long-term costs associated with nuclear waste management. In our previous study, we demonstrated that using gibbsite, instead of kyanite, as the alumina source in low-activity nuclear waste melter feeds increased the Re retention by up to 20 %, a result attributed to the formation of nanocrystalline alumina in the gibbsite-containing melter feeds. In this work, we show that using boehmite, another Al-source that forms nanocrystalline alumina polymorphs upon heating, results in an even higher retention of Re, by up to 25 %. Using X-ray diffraction and deionized water leaching tests on heat-treated feed samples, we analyzed the compositions of soluble oxyanionic salt phase and the insoluble glass-forming phase as a function of temperature, and evaluated the rate of Re incorporation into the developing transient alkali-alumino-borosilicate melt. We found that the formation of nanocrystalline alumina in the boehmite-containing feed occurred at temperatures approximately 100 °C lower than in the gibbsite feed, accelerating reactions between alumina, oxyanionic salts, and the borosilicate melt, which likely accounts for the enhanced Re retention in the boehmite-containing feeds.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155916"},"PeriodicalIF":2.8,"publicationDate":"2025-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138347","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":"Heat-treated additively manufactured and wrought 316L steels display a comparable response to ion irradiation","authors":"Benjamin M. Jenkins ,&nbsp;Solène Rouland ,&nbsp;Auriane Etienne ,&nbsp;Anna Kareer ,&nbsp;Jack Haley ,&nbsp;Cristelle Pareige ,&nbsp;Philippe Pareige ,&nbsp;Bertrand Radiguet","doi":"10.1016/j.jnucmat.2025.155913","DOIUrl":"10.1016/j.jnucmat.2025.155913","url":null,"abstract":"<div><div>Additive manufacturing produces metallic components with inhomogeneous microstructures. This inhomogeneity can negatively impact mechanical properties and in-service performance. Applying post-printing heat treatments can reduce microstructural inhomogeneity but a validation of alloy performance, under specific operational environments is still required.</div><div>316L stainless steels are used for a variety of components in nuclear power plants. They are exposed to irradiation at elevated temperature during service, which alters the microstructure and mechanical properties. To validate the implementation of additively manufactured 316L components in environments where they are exposed to irradiation, it is necessary to ensure that additively manufactured components will display comparable behaviour under irradiation to their wrought counterparts.</div><div>In this study we use atom probe tomography, transmission electron microscopy, and nanoindentation to investigate the response of additively manufactured 316L alloys, produced by laser powder bed fusion, exposed to ion irradiation. Our results, when compared to published data on wrought 316L alloys, demonstrate that performing post-printing heat treatments at 1066 °C and 1150 °C leads to 316L alloys that display a comparable response to ion irradiation when compared to conventionally manufactured 316L specimens.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155913"},"PeriodicalIF":2.8,"publicationDate":"2025-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144146731","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":"Homogeneous U0.89Pu0.11O2 mixed oxide by use of PuO2 nanopowders","authors":"Jacobus Boshoven ,&nbsp;Abibatou Ndiaye ,&nbsp;Bertrand Morel ,&nbsp;Marco Cologna ,&nbsp;Jean-François Vigier ,&nbsp;Ramon Carlos Marquez ,&nbsp;Rudy J.M. Konings ,&nbsp;Karin Popa","doi":"10.1016/j.jnucmat.2025.155915","DOIUrl":"10.1016/j.jnucmat.2025.155915","url":null,"abstract":"<div><div>We show that a homogeneous single-phase mixed oxide is obtained by blending and sintering commercial UO₂ and nanocrystalline PuO₂ powder; whereas, under the same conditions, conventional PuO₂ powder obtained from oxalate precipitation and composed of platelets, yields two-phase mixed oxide. The use of nanometric PuO₂ results in a larger final grain size.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155915"},"PeriodicalIF":2.8,"publicationDate":"2025-05-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138346","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 characterization of two variants of fuel cladding chemical interaction in a HT-9 Clad U-10Zr Fuel. Variant 1: FCCI with a Zr Rind","authors":"Daniele Salvato ,&nbsp;Bao-Phong Nguyen ,&nbsp;Yachun Wang ,&nbsp;Fidelma Giulia Di Lemma ,&nbsp;Luca Capriotti ,&nbsp;Assel Aitkaliyeva ,&nbsp;Tiankai Yao","doi":"10.1016/j.jnucmat.2025.155855","DOIUrl":"10.1016/j.jnucmat.2025.155855","url":null,"abstract":"<div><div>This study investigated the fuel cladding chemical interaction (FCCI), a key factor that limits operational temperature and burnup, in an HT-9 clad U-10Zr nuclear fuel sample irradiated to a high burnup of 13.1 at.% at a time-averaged peak inner cladding temperature (PICT) of 530 °C. Previous results showed this fuel sample exhibited two distinct levels of FCCI at d. This paper analyzed the FCCI at an azimuthal position showing an interdiffusion layer of &lt;10 µm using transmission electron microscopy to examine chemical and crystallographic nature of phases at the fuel-cladding interface at the nanoscale level. A ZrC layer and a Zr₃Si phase were identified at the interface; these, along with the relatively low local temperature, potentially contributed to limit interdiffusion, behaving as inhibitors for deleterious interactions. Lanthanides (Ln) partially consumed the ZrC layer and interacted with Fe, forming a Zr-Ln compound and a (Zr,Ce)Fe₂₊<em>_x</em> phase while also infiltrating up to 4 µm into the cladding. Neither U nor Zr were observed in the cladding, whereas Fe diffused up to 3–5 µm in the fuel. Fe infiltration formed a ternary U-Zr-Fe ε-phase and likely promoted the precipitation of a Cr-rich α’ phase on the cladding interface. Additionally, a Cr-rich χ-phase, likely formed by the dissociation of pre-existing M₂₃C₆ carbide precipitates, was identified about 2–5 µm from the fuel-cladding interface. Irradiation-induced nano-voids were also observed in the HT-9 bulk. These findings provide critical insights into FCCI mechanisms at representative irradiation conditions, essential for developing models simulating in-pile metallic fuel behaviors for next-generation reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155855"},"PeriodicalIF":2.8,"publicationDate":"2025-05-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144099411","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":"Effect of gamma radiation on early age strength and pore structure development of metakaolin-based geopolymer used for conditioning cesium and strontium radioactive waste","authors":"Emile Mukiza ,&nbsp;Quoc Tri Phung ,&nbsp;Suresh Seetharam ,&nbsp;Lander Frederickx ,&nbsp;Ken Verguts ,&nbsp;Eef Weetjens ,&nbsp;Manu K. Mohan ,&nbsp;Geert De Schutter","doi":"10.1016/j.jnucmat.2025.155912","DOIUrl":"10.1016/j.jnucmat.2025.155912","url":null,"abstract":"<div><div>This paper presents the effect of gamma radiation on early age strength and pore structure development in metakaolin (MK)-based geopolymers containing realistic cesium and strontium loading determined based on Boom Clay, a hypothetical host formation under consideration in the Belgian geological disposal concept. The effect of gamma-induced heat was decoupled from the ionizing nature of gamma radiation and assessed separately to elucidate its impact on strength and microstructure development. Changes in pore structure were evaluated using nitrogen physisorption and mercury intrusion porosimetry. The results indicate that both gamma radiation and temperature, analogous to irradiation-induced heat, negatively impacted the strength and pore structure development. Gamma radiation exposure of fresh geopolymer samples resulted in a coarser microstructure, leading to lower strength. No dose rate effect was observed, but the type of gamma radiation source had a significant impact, particularly on pore structure. Geopolymer samples exposed to Cs-137 from spent nuclear fuel exhibited larger pore structure alteration than those exposed to Co-60 at the same cumulative dose and similar dose rates. This suggests that a higher degree of pore structure alteration than previously reported in the literature could be anticipated in real-world Cs and Sr immobilization. The pore structure alteration is attributed to both gamma-induced heat and gamma-assisted water radiolysis and subsequent H₂ evolution and escape, with water radiolysis being the dominant mechanism of microstructural damage. Nevertheless, the MK-based geopolymer exposed to gamma radiation during hardening maintained satisfactory compressive strength, demonstrating strong radiation resistance. This indicates that MK-based geopolymer is promising for the immobilization of Cs and Sr-containing wastes. This study not only provides insights on formulating waste forms with realistic waste content in line with the foreseen geological disposal concept, but also advances our knowledge on mechanical and pore structure development under gamma irradiation, which have positive implications on radioactive waste management.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155912"},"PeriodicalIF":2.8,"publicationDate":"2025-05-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144115707","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 impact of chromium coatings on Zircaloy cladding deformation behavior under reactivity-initiated accident-like mechanical loading conditions","authors":"Jennifer I. Espersen ,&nbsp;Ben E. Garrison ,&nbsp;Petr Cervenka ,&nbsp;Arunkumar Seshadri ,&nbsp;Kory Linton ,&nbsp;Koroush Shirvan ,&nbsp;Nathan A. Capps ,&nbsp;Nicholas R. Brown","doi":"10.1016/j.jnucmat.2025.155910","DOIUrl":"10.1016/j.jnucmat.2025.155910","url":null,"abstract":"<div><div>A reactivity-initiated accident (RIA) occurs when a control rod ejection or control blade drop causes an increase in the fission rate. The injection of energy results in an increase in fuel temperature which in turn causes rapid thermal expansion of the fuel pellet. This thermal expansion may result in pellet-cladding mechanical interaction (PCMI) in which the fuel imparts a mechanical strain to the cladding. PCMI may cause the cladding to fail, and thus, the mechanical response of cladding due to PCMI must be investigated when characterizing new cladding materials. Chromium-coated Zircaloy-4 is a near-term accident-tolerant fuel cladding that exhibits improved high-temperature oxidation resistance. Modified burst testing was utilized to experimentally simulate the effects of PCMI on both uncoated and chromium-coated Zircaloy cladding samples at hot zero power conditions. Samples were coated using either cold spraying or physical vapor deposition to understand the differences in behavior that the coating application method may cause. Digital image correlation was used to analyze images of the deforming specimens to extract the in-situ strain behavior of the cladding. The uncoated specimens burst at hoop strains ranging from 8.8 % to 17.2 %. The cold-spray chromium-coated Zircaloy specimens burst at hoop strains of 7.0 % to 11.0 %. The physical vapor deposition coated tubes burst at hoop strains of 9.1 % to 11.5 %. These results indicate that the chromium coating causes a loss in the ductility of the cladding. The higher burst hoop strains of the physical vapor deposition-coated samples relative to the cold-spray samples indicate that the cold-spraying technique causes more of a loss in ductility than physical vapor deposition. All samples burst at higher hoop strains than those expected to occur in an RIA for fresh cladding.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155910"},"PeriodicalIF":2.8,"publicationDate":"2025-05-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144115706","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 IGR research reactor’s uranium-graphite fuel's high-temperature corrosion by a combination of thermal analysis and mass-spectrometry methods","authors":"Kuanysh Samarkhanov,&nbsp;Yuriy Ponkratov,&nbsp;Timur Kulsartov,&nbsp;Yuliya Baklanova,&nbsp;Yuriy Gordienko,&nbsp;Yerzhan Sapatayev,&nbsp;Vadim Bochkov","doi":"10.1016/j.jnucmat.2025.155908","DOIUrl":"10.1016/j.jnucmat.2025.155908","url":null,"abstract":"<div><div>The National Nuclear Center of the Republic of Kazakhstan (NNC RK) operates two unique research reactors – IVG.1 M and IGR, both of which are undergoing a conversion program to reduce nuclear fuel enrichment from 90 % to 19.75 % by ²³⁵U In 2023, the conversion of the IVG.1 M reactor was successfully completed. NNC RK is currently conducting various studies related to the conversion to a new uranium-graphite fuel for the IGR reactor. One such study focuses on the corrosion processes of the fuel surface under normal operating conditions and in various emergency situations, associated with the penetration of oxygen and water vapor into the reactor core. The study of changes in the basic physical and chemical properties of uranium-graphite fuel during interaction with chemically active gases and vapor-gas mixtures is crucial for predicting material behavior under various operating conditions.</div><div>This paper presents the results of the study of high-temperature corrosion of unirradiated uranium-graphite fuel of the IGR research reactor, using a combination of thermal analysis and mass spectrometry methods. During the experiments, data were obtained on the sample temperature, sample mass, heat flux to the sample, and changes in the composition of the reaction gas (vapor-gas mixture) in the chamber, which was purged throughout the experiment. The results of the high-temperature corrosion experiments of experimental, highly enriched uranium-graphite fuel in the reaction chamber of a thermogravimetric analyzer, with different compositions of chemically active gases, are presented. An analysis of the experimental data enabled the identification of corrosion mechanisms and the determination of the parameters of these processes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155908"},"PeriodicalIF":2.8,"publicationDate":"2025-05-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144105083","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 kinetic mechanism of dehydrogenation based on yttrium hydride surfaces: First principles calculations","authors":"Meng Tang ,&nbsp;Xi Wang ,&nbsp;Xunxiang Hu ,&nbsp;Guangdong Liu ,&nbsp;Zhixiao Liu ,&nbsp;Huiqiu Deng","doi":"10.1016/j.jnucmat.2025.155903","DOIUrl":"10.1016/j.jnucmat.2025.155903","url":null,"abstract":"<div><div>A fundamental understanding of the dehydrogenation mechanisms of yttrium hydride (YH₂) is of great importance for its applications in neutron moderator. In this study, we have employed density-functional theory (DFT) to systematically investigate the surface structures and the relative stabilities of low Miller-index YH₂ facets, with particular emphasis on their dependence on environmental parameters (hydrogen partial pressure <span><math><mrow><msub><mi>P</mi><msub><mi>H</mi><mn>2</mn></msub></msub><mspace></mspace></mrow></math></span>and temperature <em>T</em>). Based on the computational results, a surface phase diagram of YH₂ was obtained for a wide <span><math><msub><mi>P</mi><msub><mi>H</mi><mn>2</mn></msub></msub></math></span> range (300 -1100 K). The YH₂ surfaces tend to be dominated by the stoichiometric (111) facet with the 111-stoi-4H termination and the surface energy of 0.78 J/m². Two additional surfaces, 100-non-2Y and 110-non-2Y1H, will be exposed at the extremely low H partial pressure condition. Then, several possible dehydrogenation pathways were explored to elucidate the dehydrogenation mechanism based on these three surfaces. The results show that the dehydrogenation process in YH₂ primarily involves the formation of interstitial hydrogen atoms within the bulk, followed by the migration of these interstitial atoms from the bulk to the surface, ultimately leading to the release of hydrogen atoms into the surrounding environment.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155903"},"PeriodicalIF":2.8,"publicationDate":"2025-05-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144115701","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}