Journal of Nuclear Materials最新文献

{"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}

{"title":"The mechanical behavior and fracture of chromium film-zirconium alloy substrate systems","authors":"T. Hasan,&nbsp;M.A. Zikry","doi":"10.1016/j.jnucmat.2025.155909","DOIUrl":"10.1016/j.jnucmat.2025.155909","url":null,"abstract":"<div><div>A microstructurally-based dislocation crystalline plasticity (DCP) approach was integrated with a fracture approach to predict and understand how fracture nucleation and propagation at different length scales affects the mechanical behavior of a thin chromium film on a zirconium alloy substrate, which is a system representative of systems used for harsh environments, such as those pertaining to nuclear cladding systems. The grain-morphologies and orientations are based on experimental observations and these morphologies in combination with the crystalline mismatches of a b.c.c. chromium thin film and an h.c.p. zirconium alloy have a significant effect on fracture nucleation and growth. The predictions were validated with experimental observations and measurements. Transgranular fracture modes first nucleated in the chromium layer and propagated to the to the substrate. These validated results indicate that chromium coatings can delay the onset of failure in zirconium substrates or claddings, but eventually the film will be degraded due to nucleated population fracture cracks.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155909"},"PeriodicalIF":2.8,"publicationDate":"2025-05-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144105082","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":"Comparison of scanned and defocussed beam ion irradiation hardening of UHP Fe and Fe-Cr alloys","authors":"Luke Hewitt ,&nbsp;Chris Hardie ,&nbsp;Steve Roberts","doi":"10.1016/j.jnucmat.2025.155906","DOIUrl":"10.1016/j.jnucmat.2025.155906","url":null,"abstract":"<div><div>Charged particle irradiation is often used as a surrogate for neutron irradiation, and some facilities use a beam focussed to spot which is then scanned over the target surface to ensure a uniform fluence and allow fine adjustment of the flux. This results in intermittent irradiation of an area segment at a higher instantaneous flux than the time-averaged value, in contrast to the quasi steady state exposure in a reactor environment. Recent experiments have shown this pulsed irradiation to influence damage microstructure in nuclear relevant structural materials, raising questions about its use to emulate in-service conditions. Here we compare hardening of ultra high purity Fe and Fe-Cr alloys irradiated with scanned and defocussed beams to investigate whether difference in instantaneous fluxes produces a dose rate effect. No significant difference in irradiation hardening was observed. This is discussed in terms of the mechanisms leading to irradiation induced microstructure, and suggests a regime exists where beam scanning may have a negligible effect on irradiation hardening.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155906"},"PeriodicalIF":2.8,"publicationDate":"2025-05-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144090292","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":"New phases in brown ceramics of Chornobyl lava","authors":"S.V. Gabielkov,&nbsp;I.V. Zhyganiuk,&nbsp;V.G. Kudlai,&nbsp;B.S. Savchenko,&nbsp;P.E. Parkhomchuk,&nbsp;S.O. Chikolovets","doi":"10.1016/j.jnucmat.2025.155893","DOIUrl":"10.1016/j.jnucmat.2025.155893","url":null,"abstract":"&lt;div&gt;&lt;div&gt;Thirty-nine years after the Chornobyl reactor accident, uranium-containing crystalline phases magnesium Zippeite &lt;span&gt;&lt;math&gt;&lt;mi&gt;Mg&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;3.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;[&lt;/mo&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;mi&gt;U&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;]&lt;/mo&gt;&lt;/math&gt;&lt;/span&gt;, sodium uranate &lt;span&gt;&lt;math&gt;&lt;mi&gt;N&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;a&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mi&gt;U&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;, and uranium-free crystalline phases: trigonal silicon oxide &lt;span&gt;&lt;math&gt;&lt;mi&gt;Si&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;, calcium silicate &lt;span&gt;&lt;math&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;a&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;3&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mi&gt;Si&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;5&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;, calcium silicate &lt;span&gt;&lt;math&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;a&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mi&gt;Si&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;, and magnesium silicate &lt;span&gt;&lt;math&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;g&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mi&gt;Si&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; were discovered for the first time in brown ceramics of lava-like fuel-containing materials (LFCM). The presence of urania &lt;span&gt;&lt;math&gt;&lt;mi&gt;U&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2.338&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;U&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;9&lt;/mn&gt;&lt;mo&gt;+&lt;/mo&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;/math&gt;&lt;/span&gt;, cubic zirconia c-&lt;span&gt;&lt;math&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;mi&gt;Z&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;r&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;1&lt;/mn&gt;&lt;mo&gt;−&lt;/mo&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;U&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;, chornobylite &lt;span&gt;&lt;math&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;mi&gt;Z&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;r&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;1&lt;/mn&gt;&lt;mo&gt;−&lt;/mo&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;U&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;mi&gt;Si&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; and iron &lt;em&gt;α&lt;/em&gt;-Fe has been confirmed. Most of the uranium in the crystalline phases is in the form of magnesium zippeite &lt;span&gt;&lt;math&gt;&lt;mi&gt;Mg&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;3.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;[&lt;/mo&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;mi&gt;U&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mn","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155893"},"PeriodicalIF":2.8,"publicationDate":"2025-05-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144084690","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":"Structure evolution and tin redistribution during oxidation of Zircaloy-4 at 500°C","authors":"Josephine Hartmann ,&nbsp;Tamas Varga ,&nbsp;Caleb Schenck ,&nbsp;Chris McRobie ,&nbsp;Fu-Yun Tsai ,&nbsp;Vaithiyalingam Shutthanandan ,&nbsp;Arun Devaraj ,&nbsp;David Senor ,&nbsp;Bharat Gwalani ,&nbsp;Elizabeth Kautz","doi":"10.1016/j.jnucmat.2025.155895","DOIUrl":"10.1016/j.jnucmat.2025.155895","url":null,"abstract":"<div><div>Zirconium (Zr) alloys are widely used as fuel cladding in nuclear power reactors due to their thermal stability, mechanical durability, corrosion resistance, and low neutron absorption cross-section. However, their performance is challenged by oxidation in reactor environments, making the study of Zr alloy corrosion behavior crucial for ensuring the safety, longevity, and economic viability of nuclear power systems. While the oxidation behavior of Zr-based cladding materials has been extensively studied since the 1950s, a mechanistic understanding into the relationship between structure evolution, solute element redistribution, and properties remains elusive. Valuable insights may be obtained through advanced experimental methods, such as in-situ and high resolution microscopy techniques. In this study, the oxidation behavior of Zircaloy-4 at 500<!--> <!-->°C in O₂ is characterized using a multimodal advanced characterization approach. Using in-situ X-ray diffraction, the phase evolution from metastable to stable oxides is tracked in real time. Complementary high-resolution techniques, including electron microscopy and atom probe tomography, reveal nanoscale insights into the microstructural changes and solute redistribution across the oxide/metal interface. Nanohardness mapping across the oxide/metal interface highlights localized mechanical property variations that may be linked to changes in microstructure and crystal structure within the oxide layer. These findings offer valuable insights into the microstructure and property evolution of Zircaloy-4 during oxidation, contributing to a better understanding of microstructural changes in Zr-based alloys under oxidative environments.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155895"},"PeriodicalIF":2.8,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144099409","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":"PWHT effect on the microstructure and microhardness of boron-added stainless steel laser weldments","authors":"Joonghoon Kim ,&nbsp;Jongmin Han ,&nbsp;Gang Ho Lee ,&nbsp;Jaeho Jang ,&nbsp;Hyoung Chan Kim ,&nbsp;Byung Jun Kim ,&nbsp;Dae-Geun Nam ,&nbsp;Jong Bae Jeon ,&nbsp;Byoungkoo Kim","doi":"10.1016/j.jnucmat.2025.155905","DOIUrl":"10.1016/j.jnucmat.2025.155905","url":null,"abstract":"<div><div>ASTM <span><span>A887</span><svg><path></path></svg></span> 304B4 borated stainless steel (BSS) is a neutron-absorbing material developed for the structures for storing spent nuclear fuel. This study investigated the microstructure, boride morphology, and hardness changes in the weldment under various post-weld heat treatment (PWHT) conditions to determine the optimal PWHT conditions for laser-welded BSS structures. 6 mm thick 1.09B BSS plates were joined using laser welding. The PWHT was performed at 400, 800, and 1200 °C for 1, 4, 24, 100, and 200 h, respectively. The microstructure and microhardness were analyzed across the weld metal (WM), heat-affected zone (HAZ), and base metal (BM). The laser welds were composed of fine dendrites and inter-dendritic boride eutectic, showing higher hardness than the BM due to the effective grain refinement. At 400 and 800 °C PWHT, no changes in the eutectic boride morphology of the welds were observed, but the hardness of the HAZ decreased to a level similar to that of the BM as the PWHT time increased. At 1200 °C, boride spheroidization occurred from 1 hour, and the dendritic structure disappeared. From 100 h, homogenization occurred in the optical-microscope microstructure, where the distinction between the WM and BM disappeared, and the hardness became uniform due to the homogenization of the boride morphology and distribution. The formation and growth of borides were simulated using Thermo-Calc and TC-PRISMA, yielding results similar to the actual microstructural changes. The microstructural changes and physical properties observed after PWHT at 1200 °C provide evidence for exemption from periodic inspections of weldments in nuclear structures.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155905"},"PeriodicalIF":2.8,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144071787","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":"Ductile-to-brittle transition behavior and mechanism of Cr-coated Zr-1Nb alloys after 1000–1200 °C steam oxidation","authors":"Jianxi Deng,&nbsp;Shangle Huo,&nbsp;Donghui Geng,&nbsp;Qiaoyan Sun,&nbsp;Zhongxiao Song,&nbsp;Jun Sun","doi":"10.1016/j.jnucmat.2025.155904","DOIUrl":"10.1016/j.jnucmat.2025.155904","url":null,"abstract":"<div><div>Mechanical properties, particularly ductility, are crucial for evaluating the safety threshold and limits of Cr-coated zirconium claddings at loss of coolant accident (LOCA). The mechanical properties of Cr-coated Zr-1Nb alloy plates after 1000–1200 °C steam oxidation are investigated through uniaxial tensile tests at 135 °C, with uncoated specimens as comparison. The results show that the ductile-to-brittle transition of Cr-coated samples is significantly affected by oxidation temperatures. After 2 h oxidation at 1000 °C, Cr-coated samples remain good ductile with 16.2 % elongation compared to 0.4 % for uncoated ones. While, Cr-coated samples exhibit brittle fracture with 0.2 % elongation after 2 h oxidation at 1150 °C, and the tensile strength diminishes to 321 MPa, a reduction of 18 % compared to before oxidation. The ductile-to-brittle transition is associated with the protective structure evolution of Cr coating, which essentially determines the formation of brittle α-Zr(O) in Zr substrate. At 1000 °C, good oxygen diffusion barrier of Cr coating significantly reduces α-Zr(O) layer formation. However, accelerated oxidation of Cr coating results in the generation of α-Zr(O) phase. The critical α-Zr(O) intrusion depth for brittle fracture is about 85 % of Zr substrate, and the content of O and Cr increased by 0.96 % (wt.) and 1.84 % (wt.), respectively. Formation of brittle α-Zr(O) dominates the ductility degradation, and O and Cr element diffusion into substrate also have a certain effect. The findings can provide insights for performance degradation modeling of Cr-coated Zr alloys under LOCA conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"614 ","pages":"Article 155904"},"PeriodicalIF":2.8,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144090293","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}