Journal of Nuclear Materials最新文献

{"title":"Microstructurally informed synchrotron x-ray analysis revealing helium defect transitions in ultrafine grained tungsten","authors":"Cormac Killeen ,&nbsp;Yang Zhang ,&nbsp;W. Streit Cunningham ,&nbsp;David J. Sprouster ,&nbsp;Daniel Olds ,&nbsp;Shirish Chodankar ,&nbsp;Jason R. Trelewicz","doi":"10.1016/j.jnucmat.2025.156106","DOIUrl":"10.1016/j.jnucmat.2025.156106","url":null,"abstract":"<div><div>The formation of insoluble gaseous defects in materials due to nuclear transmutation or ion implantation involves the diffusion of impurity atoms to form atomic defect clusters that coalesce into bubbles or cavities and ultimately degrade the material properties. Transmission electron microscopy (TEM) is limited in its ability to resolve sub-nanometer gas clusters whereas X-ray diffraction (XRD) provides information pertaining to local atomic changes. In this study, helium (He) implanted ultrafine grained tungsten is explored through a multimodal defect characterization campaign combining TEM-informed Small Angle X-ray Scattering (SAXS) analysis, XRD lattice parameter measurements, and nanoscale He cluster quantification from a region of reciprocal space accessible via Wide Angle X-ray Scattering (WAXS). Moderate elevated temperature implantations are shown to produce high concentrations of sub-nanoscale He clusters and small, homogeneously distributed cavities, which collectively are linked to lattice expansion and further substantiated through complementary atomistic simulations. Increased implantation temperatures encourage the diffusion of these defects to the grain boundaries (GBs), leading to lattice relaxation and the growth of large GB cavities manifesting as bimodal size distributions in the SAXS analysis. Overall, our results demonstrate the utility of multimodal synchrotron X-ray analysis in bridging the gap between microscale He cavity quantification and atomic-scale defect analysis.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156106"},"PeriodicalIF":3.2,"publicationDate":"2025-08-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144886997","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 effect and damage mechanism of 5.3 MeV α-irradiation on O-methyl phenolic epoxy resin","authors":"Zhao Sun ,&nbsp;Yunyun Lu ,&nbsp;Liyuan Deng ,&nbsp;Zhencen He ,&nbsp;Qi Cao ,&nbsp;Zhimin Hu","doi":"10.1016/j.jnucmat.2025.156089","DOIUrl":"10.1016/j.jnucmat.2025.156089","url":null,"abstract":"<div><div>This study elucidated the structural damage mechanisms of O-methyl phenolic epoxy resin (EOCN) under <em>α</em>-irradiation up to 6 MGy. In situ nanogram-level mass measurements using quartz crystal microbalance with dissipation (QCM-D) revealed accelerated degradation beginning at 2 MGy, accompanied by persistent relaxation effects. Fourier-transform infrared spectroscopy (FTIR) and X-ray photoelectron spectroscopy (XPS) analyses confirmed that the cleavage of epoxy groups constituted the primary degradation pathway, while the simultaneous scission of multiple functional groups contributed to a complex degradation mechanism. Electron paramagnetic resonance (EPR) spectroscopy demonstrated a dose-dependent increase in the concentration of singlet free radicals. Quantum chemical calculations and reactive force field molecular dynamics (ReaxFF-MD) simulations were employed to qualitatively analyze the physicochemical characteristics of key functional groups identified experimentally. Based on these results, two ether bond-breakage degradation pathways were induced by <em>α</em>-particles were reconstructed, along with the associated evolution processes of small-molecule degradation products.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156089"},"PeriodicalIF":3.2,"publicationDate":"2025-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144842643","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":"Review of hydrogen-induced mechanical degradation in uranium and uranium alloys","authors":"Mengsheng Zhai,&nbsp;Sheng Zhang,&nbsp;Wenliang Xu,&nbsp;Hefei Ji,&nbsp;Lizhu Luo,&nbsp;Kunming Yang,&nbsp;Shushan Cui,&nbsp;Shilv Yu,&nbsp;Chuan Mo,&nbsp;Ruiwen Li,&nbsp;Dongli Zou,&nbsp;Dawu Xiao,&nbsp;Bin Su,&nbsp;Wenhua Luo","doi":"10.1016/j.jnucmat.2025.156102","DOIUrl":"10.1016/j.jnucmat.2025.156102","url":null,"abstract":"<div><div>In this review, we examine the hydrogen corrosion behavior of uranium and its alloys, with a focus on its impact on their mechanical performance, including hydrogen embrittlement and hydrogen-induced stress corrosion cracking (HISCC). Alloying with elements such as Nb and Mo has been shown to improve hydrogen corrosion resistance of uranium alloys, thereby reducing susceptibility to hydrogen embrittlement and hydrogen-induced stress corrosion cracking. Although studies about uranium hydrogen embrittlement under high strain deformation remain limited, existing evidence suggests that increasing strain rate can alleviate hydrogen embrittlement. The mechanisms underlying hydrogen embrittlement and hydrogen-induced stress corrosion cracking in uranium are still under debate, particularly regarding whether uranium hydrides formation or solid-solution hydrogen plays the dominant role in the embrittlement process. Future research should integrate advanced characterization techniques with multiscale modelling, spanning from atomic scale to continuum scale, to elucidate these mechanisms.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156102"},"PeriodicalIF":3.2,"publicationDate":"2025-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144842103","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":"An azimuthal contact and free-centerline boundary condition for cracked fuel pellets in 2D axisymmetric models","authors":"Andrew Prudil","doi":"10.1016/j.jnucmat.2025.156101","DOIUrl":"10.1016/j.jnucmat.2025.156101","url":null,"abstract":"<div><div>A fuel pellet deformation model comprised of plane-stress, azimuthal contact, and free-centerline displacement boundary condition is proposed to improve the prediction of Pellet-Cladding Mechanical Interaction (PCMI) in two-dimensional (2D) axisymmetric fuel codes at low power. This modification is motivated by case 3 of the EGRFP-PCMI modelling benchmark, in which none of the participants consistently correctly predicted the linear power at which pellet-cladding contact starts, nor the relationship of this threshold power with pellet length. In the benchmark, the 2D fuel codes with discrete fuel pellets did not significantly outperform the one-dimensional codes in predicting PCMI and the effects of pellet length. In the present work, we analyze the predictions of the Canadian Fuel And Sheath modelling Tool (CFAST) in the benchmark study and argue that the zero radial displacement solid-mechanics boundary condition commonly used in 2D axisymmetric models limits pellet hourglassing and is inconsistent with the expected pellet cracking. This hypothesis is confirmed by a three-dimensional (3D) thermomechanical model with cracked fuel pellets compared against experimental measurements of clad elongation. Based on these findings, the modified 2D axisymmetric model is proposed and compared against the 3D model and the experimental measurements of cladding elongation at low power.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156101"},"PeriodicalIF":3.2,"publicationDate":"2025-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144842104","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":"Elucidating the effect of minor-actinide addition on fuel-cladding chemical interaction in an HT-9 clad U-Pu-Zr metallic fuel irradiated to 6.15 at.% burnup in EBR-II","authors":"Bao-Phong Nguyen ,&nbsp;Luca Capriotti ,&nbsp;Assel Aitkaliyeva ,&nbsp;Yachun Wang","doi":"10.1016/j.jnucmat.2025.156100","DOIUrl":"10.1016/j.jnucmat.2025.156100","url":null,"abstract":"<div><div>Scanning and transmission electron microscopy (S/TEM) were used to characterize the local fuel-cladding chemical interaction (FCCI) in one cross-section taken from a HT-9 clad U-20.3Pu-10Zr-1.2Am-1.3Np (in wt.%) fuel irradiated to 6.15 at.% burnup with inner cladding temperatures ranging between 460–490 °C. Results showed that the total interaction thickness between fuel and cladding was &lt;10 µm. Fe infiltrated the fuel to form U-Zr-Fe phases while fuel elements or lanthanides did not infiltrate into the cladding. Np was not involved in the formation of any phases in the examined locations; however, Am played a role by forming a ∼2 µm thick homogeneous Fe-Pu-Am planar front at the inner cladding wall. An oxidized Na layer existed in the fuel-cladding gap with Fe and lanthanide particles dispersed within, suggesting Na could facilitate the transport of fuel and cladding constituents. Secondary phases, including an FCC Zr-rich phase, lanthanide phases, and α’-Cr(Fe) were identified in the outer fuel and FCCI regions. This study suggests that, for the irradiation conditions specific to this cross-section, minor actinides have little impact on FCCI behavior beyond what would be observed in typical HT-9 clad U-Pu-Zr fuel pins systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156100"},"PeriodicalIF":3.2,"publicationDate":"2025-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144878877","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":"Helium-to-vacancy ratio of helium bubbles in RAFM steels irradiated in STIP: A new perspective via MD simulation","authors":"Yeshang Hu ,&nbsp;Lei Peng ,&nbsp;Jingyi Shi ,&nbsp;Yiyi Ma ,&nbsp;Yao Xie ,&nbsp;Zhenyu Wei ,&nbsp;Yongjie Sun ,&nbsp;Yuanxi Wan","doi":"10.1016/j.jnucmat.2025.156096","DOIUrl":"10.1016/j.jnucmat.2025.156096","url":null,"abstract":"<div><div>The helium-to-vacancy (He/V) ratio plays a crucial role in the helium bubble-induced damage mechanisms caused by neutron irradiation in reduced activation ferritic/martensitic (RAFM) steels, which are main candidate structural materials for fusion reactors. Based on the results of hardening induced by helium bubble in RAFM steel specimens irradiated in the Swiss spallation neutron source, molecular dynamics (MD) simulations were conducted to investigate the interaction between edge dislocation and helium bubble with varying sizes and He/V ratios. The barrier strength of helium bubbles were calculated based on the dispersed barrier hardening model. From a new perspective, the He/V ratio of helium bubbles in STIP specimens was evaluated through comparing the barrier strength obtained from experimental hardening data and MD simulations. The results showed that the barrier strength of bubbles initially increased slightly as the He/V ratio increased from 0 to 0.8, reached its peak within the He/V ratio range of 0.8 to 1.1, and then decreased rapidly to a very low level. By comparing the simulated and experimental barrier strength, the He/V ratio range of bubbles in RAFM steel specimens was estimated. The He/V ratio of He bubbles in RAFM steel specimens with middle doses is found to be within the high He/V ratio range of 1.2–1.4, whereas those with low doses and a high dose are situated within the He/V ratio range of 0.8–1.1. Furthermore, the He/V ratio of the helium bubbles in RAFM steels was analyzed in conjunction with existing results.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156096"},"PeriodicalIF":3.2,"publicationDate":"2025-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144842644","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":"Influence of symmetric tilt grain boundaries and/or Cr-rich α′ precipitates on irradiation damage in Fe-Cr-Al alloys: A molecular dynamics investigation","authors":"Danting Zheng ,&nbsp;Dong Chen ,&nbsp;Haosheng Zhao ,&nbsp;Lizhi Li ,&nbsp;Zhiheng Guo ,&nbsp;Zuncheng Zhao ,&nbsp;Huiqiu Deng","doi":"10.1016/j.jnucmat.2025.156090","DOIUrl":"10.1016/j.jnucmat.2025.156090","url":null,"abstract":"<div><div>Recently, Fe-Cr-Al alloy is considered one of the most prospective accident-tolerant fuel cladding materials. However, Fe-Cr-Al alloy exposed to fast neutron environments exhibits microstructural features, which leads to material embrittlement, decreases mechanical properties, and affects its service performance. In the present work, molecular dynamics (MD) methods are employed to simulate the irradiation performance of three Fe-Cr-Al alloy systems, namely alloys containing grain boundary (alloy-GB systems), alloys containing Cr-rich α′ phase (alloy-α′ systems), and alloys containing GB and Cr-rich α′ phase (alloy-GB-α′ systems). The GB energies (<em>E</em>_GB) for different symmetric tilted grain boundaries (STGBs) are calculated. The STGBs with high <em>E</em>_GB exhibit more point defects and lower defect annihilation rate than those with low <em>E</em>_GB. And the defect annihilation rate of STGB with low <em>E</em>_GB is high with fewer defects surviving in the bulk. Moreover, the number of vacancies in the bulk is always greater than the number of interstitials. For the alloy-α′ systems containing the Cr-rich α′ phase with the [001] orientation, the increases in temperature and the distance of PKA from the center of the Cr-rich α′ phase (<em>d</em>_PKA-α′) will reduce the degree of irradiation damage. The increases in the size and the Cr content of the Cr-rich α′ phase will enhance the degree of irradiation damage. For the alloy-GB-α′ systems with Σ19(116)[110] GB, the number of final surviving defects is greater than that in the alloy-GB systems, which indicates that the interaction of GB and Cr-rich α′ phase will further increase the degree of irradiation damage. The increases of <em>d</em>_PKA-GB will deepen the degree of irradiation damage when the PKA atom is located in/beyond the Cr-rich α′ phase. Moreover, the dislocations form on the GB for the alloy-GB-α′ systems with Σ19(116)[110]. The appearance of dislocation can significantly affect the defect behaviors in the irradiation damage, which results in a decrease in the number of defects and defect clustering rate.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156090"},"PeriodicalIF":3.2,"publicationDate":"2025-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144842645","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":"An integrated approach to examine fuel-cladding chemical interaction in HT9/U-10Zr metallic fast reactor fuels: Coupling machine learning with electron microscopy and local mechanical properties analysis","authors":"Luca Capriotti,&nbsp;Fidelma Di Lemma,&nbsp;Daniele Salvato,&nbsp;Fei Xu,&nbsp;Yalei Tang,&nbsp;Kyle M. Paaren,&nbsp;Alexander L. Swearingen,&nbsp;Colby B. Jensen,&nbsp;Yachun Wang,&nbsp;Douglas L. Porter","doi":"10.1016/j.jnucmat.2025.156092","DOIUrl":"10.1016/j.jnucmat.2025.156092","url":null,"abstract":"<div><div>The metallic U-Zr nuclear fuel alloy has garnered renewed interest as a promising candidate for next-generation sodium-cooled fast reactors. Recent studies and technology assessments have identified several areas requiring improvements, enhanced knowledge, and reliable data to strengthen the U-Zr fuel design basis for qualification and commercial applications. One of the most challenging phenomena impacting this fuel system’s performance is fuel-cladding chemical interaction (FCCI). This work aimed to harvest FCCI data by examining selected HT9/U-10Zr (wt. %) fuel samples of prototypic full-length fuel pins through an integrated approach. This approach integrated scanning electron microscopy (SEM) microstructure characterization with localized mechanical properties examination to deepen understanding of FCCI phenomenon in HT9/U-10Zr fuel system. Particularly, this study focused on MFF fuel pins irradiated at Fast Flux Test Facility (FFTF), which aimed to qualify metallic fuel as a driver fuel for FFTF and to assess its viability for larger-scale fast reactors. Electron microscopy provided high confidence in detecting and distinguishing the different FCCI layers, while small-scale mechanical testing (SSMT) probed the mechanical properties of these layers. SEM examination of a MFF-2 pin 192167, with a time averaged inner cladding temperature (TICT) slightly over 500°C, revealed minimal cladding-side FCCI (cladding wastage). In contrast, significantly thicker cladding wastage comprising two distinct sublayers was observed in samples from the thermally hot MFF-3 pin 193045 and MFF-5 pin 195011 where the TICT ranged from 610-635°C. SSMT indicated complete embrittlement in the sublayer adjacent to the fuel and a tendency toward embrittlement in the other sublayer. Additionally, a new machine learning method was developed, validated, and used to quantify cladding wastage thickness. The machine learning method reliably predicted the wastage thickness across various fuel pins and sample cross-sections. The available cladding wastage data from HT9/U-10Zr fuel system demonstrated a strong temperature dependency. However, the dataset remains small, and ongoing research activities are essential to further understand the FCCI phenomenon and develop a reliable FCCI model for enhanced fuel performance simulation under various conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156092"},"PeriodicalIF":3.2,"publicationDate":"2025-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144878875","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":"Characterization of in-situ and ex-situ ion-irradiated additively manufactured 316L and 316H stainless steels","authors":"Wei-Ying Chen ,&nbsp;Stephen Taller ,&nbsp;Andrea M. Jokisaari ,&nbsp;Yiren Chen ,&nbsp;Rongjie Song ,&nbsp;Xuan Zhang ,&nbsp;Lin Gao ,&nbsp;Peter M. Baldo ,&nbsp;Dzmitry Habaruk ,&nbsp;Meimei Li","doi":"10.1016/j.jnucmat.2025.156044","DOIUrl":"10.1016/j.jnucmat.2025.156044","url":null,"abstract":"<div><div>Additively manufactured (AM) 316 stainless steel (SS) differs from its wrought counterpart in its unique dislocation cell structure and the presence of segregation and oxide particles at the cell walls. This work investigated the evolution of the microstructure in laser powder bed fusion (LPBF) 316L and 316H SS under <em>in-situ</em> 1 MeV Kr ion irradiation at 600<!--> <!-->°C to 5 dpa, and <em>ex-situ</em> 4 MeV Ni ion irradiation at 300<!--> <!-->°C and 600<!--> <!-->°C from 0.2 dpa to 10 dpa, with a dose rate for all experiments of 10^-3 dpa/s. The results reveal that the dislocation cell structure results in heterogeneous formation of dislocation loops and voids, particularly at 600<!--> <!-->°C, where loops tend to form within the cell interiors while voids form at the cell boundaries. LPBF 316H has a reduced level of swelling compared to LPBF 316L due to prolonged incubation. Energy Dispersive X-ray Spectroscopy (EDS) mapping indicates Ni and Si segregation at void surfaces due to radiation-induced segregation. At 300<!--> <!-->°C, where voids are absent, the distribution of dislocation loops and stacking fault tetrahedra appears to be uniform. Dislocation cell structures mostly disappeared by 2 dpa for all conditions in this work. M₂₃C₆ carbides were observed in LPBF 316H at 600<!--> <!-->°C as early as 0.2 dpa, but not in LPBF 316L. Nanoindentation was performed to obtain the hardness of irradiated materials. This work illustrated the influence of additive manufacturing processes on microstructure evolution under irradiation, revealing the differences as well as the similarities as compared with wrought 316 SS, and the AM-related phenomenon that can potentially occur under neutron irradiation in nuclear reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156044"},"PeriodicalIF":3.2,"publicationDate":"2025-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144865592","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 formation and persistence of α-UH3 in uranium hydride under conditions relevant to metallic fuel and nuclear waste storage","authors":"Nick Hodge ,&nbsp;Hugh Godfrey ,&nbsp;Simon Everall ,&nbsp;Sarah May ,&nbsp;Andrew Diggle ,&nbsp;Matthew O’Sullivan ,&nbsp;Anna Adamska","doi":"10.1016/j.jnucmat.2025.156097","DOIUrl":"10.1016/j.jnucmat.2025.156097","url":null,"abstract":"<div><div>The formation of uranium hydride is recognised as a hazard during the storage of uranium metal owing to its potentially pyrophoric properties. Uranium hydride exists mainly in the form of cubic trihydride with two crystal structures. Denoted as α-UH₃ and β-UH₃, the two forms of hydride have different chemical properties, with α-UH₃ being the more chemically reactive phase. The formation and persistence of α-UH₃ under conditions relevant to waste storage will likely have a significant bearing on the reactivity of the residual uranium hydride in waste stored in legacy storage facilities. This study has assessed the formation and persistence of α-UH₃ at a range of temperatures. The work has shown that the fraction of α-UH₃ in uranium hydride gradually increases at decreasing formation temperatures. This means that it could potentially be the dominant phase formed under typical waste storage conditions, which, for the purposes of this paper, can be broadly considered to be typically about 25 °C, with uranium containing wastes either submerged under water, dry or drying with air access or damp in sealed containers. Therefore, in some cases, waste may be stored under conditions where a hydrogen atmosphere may be present to some extent. The work has further shown that in the absence of appreciable oxidant, α-UH₃ is stable at 30 °C and 50 °C for at least 100 d and for over 300 d at ambient temperature and pressure. Given the high α-UH₃ fraction at low formation temperatures and the stability of α-UH₃ at low storage temperatures, it is feasible that α-UH₃ could be the dominant phase in any uranium hydride remaining in uranium metal bearing waste. Hence, the properties of α-UH₃ could have a significant bearing on the behaviour of the waste during the planned retrieval and packaging of legacy uranium bearing wastes. These results add significantly to the understanding of uranium hydride behaviour in the context of storage, retrieval and packaging of uranium metal bearing wastes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156097"},"PeriodicalIF":3.2,"publicationDate":"2025-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144887117","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}