Journal of Nuclear Materials最新文献

{"title":"Abnormal growth of loops by repulsive interaction between loops and dislocations","authors":"Shen Li ,&nbsp;Yipeng Li ,&nbsp;Ziqi Cao ,&nbsp;Yifan Ding ,&nbsp;Xiaoyong Wu ,&nbsp;Ruiqian Zhang ,&nbsp;Guang Ran","doi":"10.1016/j.jnucmat.2024.155589","DOIUrl":"10.1016/j.jnucmat.2024.155589","url":null,"abstract":"<div><div>Pre-existing dislocations have a significant impact on the evolution of radiation defects, which in turn affect the mechanical properties of materials after irradiation. However, systematic research into the question of how pre-existing dislocations affect the evolution of defects is essential and has yet to be carried out. In this work, the effect of pre-existing dislocations on loop evolution in FeCrAl alloys is investigated by <em>in-situ</em> transmission electron microscopy (TEM) using Fe⁺+He⁺+H₂⁺ tri-beam irradiation. It is found that the dislocations exhibit a strong sink effect and the values of the number density and size of the dislocation loops are lower in regions with pre-existing dislocations. As the calculation of sink strength shows, the sink effect of dislocations can be weakened due to the initiation of dislocation gliding, by the absorption of point defects and by climbing. And this impaired sink effect is the main reasons for the formation of a V-shaped density profile. A coexisting repulsive interaction between loops and dislocations has been proposed, which also has an important effect on the evolution morphology of loops, and its validity has been demonstrated by the interaction energy calculations. The repulsive interaction of dislocations pushes the defects away, leading to a higher probability of nucleation and coalescence. This can lead to a greater number and abnormally rapid growth of loops. In addition, a stronger surface effect is found for the thinner foil, where the density is lower and the size is smaller. The finger-shaped loops are found to be absent in the thinner samples due to the annihilation of the large loops.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155589"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171599","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":"Mechanistic nuclear fuel performance modeling of uranium nitride","authors":"Jason T. Rizk ,&nbsp;Michael W.D. Cooper ,&nbsp;Pierre-Clément A. Simon ,&nbsp;Anton J. Schneider ,&nbsp;David A. Andersson ,&nbsp;Stephen R. Novascone ,&nbsp;Christopher Matthews","doi":"10.1016/j.jnucmat.2024.155604","DOIUrl":"10.1016/j.jnucmat.2024.155604","url":null,"abstract":"<div><div>Uranium mononitride (UN) is a nuclear fuel candidate for advanced reactor designs and an alternative being considered for light water reactors due to its higher thermal conductivity and uranium density than UO₂. As with any nuclear fuel, swelling and fission gas release are important factors for safety, while also being some of the hardest phenomena to predict with a high degree of confidence. Getting a grasp on the gas swelling behavior and release is crucial to lower the barrier for UN utilization. An accelerated swelling rate at high temperatures observed experimentally, sometimes referred to as “breakaway swelling,” further complicates the prediction of fuel performance of UN. A mechanistic model has been developed using a multiscale approach to describe the intragranular and intergranular fission gas behavior. Lower-length-scale calculations have been employed to inform models of the gas and self-diffusion behavior, resolution rate, and bubble shape. Leveraging previous work on high burnup UO₂, two populations of intragranular bubbles are considered; small bulk bubbles and larger bubbles located along dislocations. The dislocation bubbles were found to be crucial to the overall swelling behavior, and the breakaway swelling transition was associated with the transition in the gas atom diffusion mechanism from an irradiation-induced athermal diffusion regime at lower temperatures to an intrinsic thermal equilibrium regime at higher temperatures, accelerating the growth of the dislocation bubbles. Similarly, the threshold for fission gas release was associated with the grain boundary vacancy diffusivity surpassing the gas atom diffusivity at sufficiently high temperatures, allowing the over-pressurized grain boundary bubble to grow in size and interconnect. Using thermo-mechanical models with the fission gas model, two integral fuel pin assessment cases were simulated. This work demonstrates the ability of a multiscale approach to accelerate the understanding of advanced fuel forms when experimental data is limited.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155604"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155302","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":"Microstructural evolution in ion irradiated cold spray Cr coated Zr-alloy","authors":"Tyler Dabney ,&nbsp;K.N. Sasidhar ,&nbsp;Evan Willing ,&nbsp;Carson Lukas ,&nbsp;Kyle Quillin ,&nbsp;Hwasung Yeom ,&nbsp;Kumar Sridharan","doi":"10.1016/j.jnucmat.2025.155652","DOIUrl":"10.1016/j.jnucmat.2025.155652","url":null,"abstract":"<div><div>Chromium coating deposited on zirconium-alloy using cold spray technology is being considered as a potential approach for increasing the accident tolerance of light water reactors. In this study, high energy Xe²⁶⁺ ion irradiation was performed on cold spray coated Cr/Zircaloy-4 system at 350 °C to investigate structural changes within the bulk Cr coating and at the Cr/Zircaloy-4 interface. High resolution structural and compositional analysis of the as-deposited coating revealed crystallographic coherence at the interface and nanometric segregation of Cr and Fe on the Zr-alloy side of the interface. Irradiation led to homogenization of Fe and the formation of a ∼20 nm thick amorphous layer at the Cr/Zircaloy-4 interface. Thermodynamic calculations revealed that both Fe dissolution and amorphization are favored due to the increased enthalpy provided by the high energy irradiation. The coatings exhibited a heterogeneous structure over multiple length scales, consisting of elongated and dynamically recrystallized ultrafine or nanocrystalline grains, high dislocation density, gradation in grain size, and interparticle boundaries. After irradiation, the interparticle boundaries were obliterated to be replaced by recrystallized grains and subtle grain growth. Consistent with these structural changes, nanoindentation testing exhibited slight softening of the coating after irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155652"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155395","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":"Neutron flux impact on rate of expansion of quartz","authors":"Ippei Maruyama ,&nbsp;Kenta Murakami ,&nbsp;Takahiro Ohkubo ,&nbsp;Shohei Sawada ,&nbsp;Osamu Kontani ,&nbsp;Takafumi Igari ,&nbsp;Masaki Kawai ,&nbsp;Junji Etoh","doi":"10.1016/j.jnucmat.2025.155631","DOIUrl":"10.1016/j.jnucmat.2025.155631","url":null,"abstract":"<div><div>The radiation-induced expansion of concrete aggregate is a significant concern in the aging management of concrete biological shielding in nuclear power plants. Understanding the sensitivity of rock-forming minerals, particularly quartz, to neutron radiation is essential in this context. In this study, we investigated the neutron irradiation effects on different types of quartz, including synthetic quartz, metachert, sandstone, and granodiorite, under irradiation temperatures ranging from 45 to 62 ℃ and the displacement damage in quartz ranging from 0.01 to 0.23 dpa. The rate of irradiation-induced expansion was determined using X-ray diffraction/Rietveld analysis, revealing a neutron flux dependency. Our findings suggest that radiation-induced relaxation or healing processes occur within quartz. To explain the observed flux-dependent and temperature-dependent radiation-induced expansion of quartz, we propose a two-phase model that considers the pristine and expanded phases. This model, based on four key variables, i.e. neutron flux, the equivalent cross-sections for phase change, and healing parameter, successfully reproduces the radiation-induced expansion behaviors observed in quartz. The model further indicates that radiation-induced relaxation, potentially linked to the diffusion of silicon (Si) or oxygen (O) within the quartz grain, plays a healing role that mitigates radiation-induced volume expansion.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155631"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143349353","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Synergistic effect of simultaneous proton irradiation and molten lead-bismuth eutectic corrosion on FeCrAlTi(Mo) coatings","authors":"Jian Yang ,&nbsp;Wei Zhang ,&nbsp;Jijun Yang","doi":"10.1016/j.jnucmat.2024.155600","DOIUrl":"10.1016/j.jnucmat.2024.155600","url":null,"abstract":"<div><div>The synergistic effect of simultaneous proton irradiation (with a total fluence of 2.0 × 10¹⁷ ions/cm²) and lead-bismuth eutectic (LBE) corrosion (at 550 °C for 46 h) on FeCrAlTi(Mo) coatings was investigated. The surface morphology, composition distribution, crystal structure of the formed oxide scale, and structural evolution of the two coatings were systematically analyzed using SEM and TEM. The corrosion results revealed the accelerated LBE corrosion behaviour of the proton-irradiated FeCrAlTi coating, which could be attributed to irradiation-induced defects leading to enhanced diffusion and irradiation-induced grain growth, resulting in the rapid formation of a protective oxide scale. However, proton irradiation decelerated the corrosion rate of the FeCrAlTiMo coating in high-temperature LBE, which could be attributed to the irradiation-induced annihilation of the free volume, which slowed down the diffusion dynamics. Whether proton irradiation accelerated or decelerated corrosion depended on the crystal structures of the FeCrAlTi and FeCrAlTiMo coatings during the corrosion process. Moreover, the novel mechanism of irradiation-decelerated corrosion provides a new concept for designing amorphous nuclear structural materials with improved synergistic corrosion and irradiation resistance.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155600"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170531","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A peridynamic model for oxidation of T91 steel in liquid lead-bismuth eutectic","authors":"Chenwen Tian ,&nbsp;Zhikun Zhou ,&nbsp;Juan Du ,&nbsp;Shuaiqi Fan ,&nbsp;Ziguang Chen","doi":"10.1016/j.jnucmat.2024.155594","DOIUrl":"10.1016/j.jnucmat.2024.155594","url":null,"abstract":"<div><div>In this paper, we present a reaction-diffusion peridynamic model for the oxidation of T91 steel in liquid lead-bismuth eutectic (LBE). By integrating oxidation kinetics with diffusion dynamics, our model is applicable to a broad range of high-temperature oxidation processes. Validation against experimental data of oxide scale growth for T91 steel in oxygen-saturated LBE at 400–550°C and up to 13,000 h confirms its accuracy. Furthermore, we demonstrate the model's enhanced accuracy relative to traditional formula-based kinetic models. The new framework is utilized to investigate the influence of temperature and oxygen concentration on oxide scale evolution, indicating that elevated oxygen concentrations and temperatures significantly accelerate oxidation. Additionally, we explore the oxidation behavior of T91 steel with a non-uniform surface and oxide scale cracks, revealing that the outer magnetite oxide scale grows more rapidly than the inner Fe-Cr spinel scale in concave regions. Our model captures the transition of the oxide scale from an irregular to a flat, uniform morphology, aligning with experimental observations. This model demonstrates considerable potential for simulating high-temperature oxidation in complex geometries, including irregular surfaces and cracks, across various environmental conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155594"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170551","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 neutron irradiation on microstructural evolution and deformation behavior of Zirconium (Zr-1Nb) alloy","authors":"Rajesh Jarugula ,&nbsp;Patricie Halodová ,&nbsp;Mariia Zimina ,&nbsp;M Sundararaman ,&nbsp;Martina Malá ,&nbsp;Jan Klouzal ,&nbsp;Martin Ševeček ,&nbsp;Josef Běláč ,&nbsp;Radomír Řeháček ,&nbsp;Stanislav Linhart","doi":"10.1016/j.jnucmat.2024.155596","DOIUrl":"10.1016/j.jnucmat.2024.155596","url":null,"abstract":"<div><div>In this study, a detailed investigation was carried out on the recrystallized Zirconium (Zr-1Nb) alloy, irradiated in the reactor core of Temelín NPP (VVER-1000) for a duration of four annual cycles within the framework of the ALVEL-ČEZ project. The main objective of the present work is to report the microstructural evolution, with a particular focus on the radiation-induced defects and elucidate the underlying deformation mechanisms. The combined use of electron backscattered diffraction and scanning transmission electron microscopy enables to obtain the statistical details of the microstructural changes in the irradiated material. To assess the deformation behavior of the irradiated Zr-1Nb alloy, tensile testing was conducted at room temperature and 350 °C and a strain rate of 2 × 10^-3 s^-1 along the axial direction of the tube. Transmission electron microscopy observations on the deformed sample revealed the presence of dislocation-free channels within the grains. Based on the stereographic trace analysis, it was determined that slip occurred through prismatic, pyramidal and basal channels. Additionally, the <span><math><mrow><mo>{</mo><mrow><mn>01</mn><mover><mn>1</mn><mo>¯</mo></mover><mn>1</mn></mrow><mo>}</mo></mrow></math></span> <span><math><mrow><mo>〈</mo><mrow><mn>0</mn><mover><mn>1</mn><mo>¯</mo></mover><mn>12</mn></mrow><mo>〉</mo></mrow></math></span> type twinning system was also found to be activated in the deformed sample.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155596"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170992","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":"Study of pellet-clad bonding using heat treatment of axially cut slices of PWR fuel rod","authors":"L. Desgranges,&nbsp;J. Sercombe,&nbsp;Y. Pontillon","doi":"10.1016/j.jnucmat.2024.155540","DOIUrl":"10.1016/j.jnucmat.2024.155540","url":null,"abstract":"<div><div>Slices of high burn-up nuclear fuel rod were prepared with or without an axial slit in the cladding or a central hole in the fuel. After identical heat treatment at 1200 °C, slices with cut cladding exhibited a wide opening of the slit while the uncut slices showed no major change of their cladding. Post-test examination evidenced that a layer made of High Burn-up Structure remained adherent to the cladding. Mechanical simulations indicated that this bonded HBS layer could be the reason for the cladding opening. These results confirmed that the HBS can be very tightly adherent to the cladding, as postulated in previous literature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155540"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171605","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 role of pressure in carburisation: Crack dynamics in Type 316H austenitic stainless steel under CO2 environments","authors":"Mariia Zimina ,&nbsp;Alexander D. Warren ,&nbsp;Lawrence Coghlan ,&nbsp;Peter J. Thomas ,&nbsp;Marc Chevalier ,&nbsp;Peter E.J. Flewitt ,&nbsp;Tomas L. Martin","doi":"10.1016/j.jnucmat.2025.155625","DOIUrl":"10.1016/j.jnucmat.2025.155625","url":null,"abstract":"<div><div>The oxidation and carburisation of steel in advanced gas-cooled nuclear reactors limit the life of key components during plant operation. To achieve a mechanistic understanding of the origin of carburisation phenomena, simulation or replication of service exposure conditions is needed. Laboratory tests are usually performed to accelerate processes such as crack initiation and growth in the material, and to simulate plant conditions as precisely as possible. This paper describes the results of carburisation data analysis of AISI Type 316H stainless steel under several CO₂ gas exposure conditions. The behaviour of ex-service welded pipe components from advanced gas-cooled reactors at service temperature and pressure is compared to a range of specimens loaded and cracked in air and under surrogate/simulated CO₂ conditions. Both atmospheric pressure and pressurised CO₂ gas (4.14 MPa) conditions were studied. Microstructural characterisation of the crack tip region and the bulk/matrix of the specimens was performed using advanced electron microscopy techniques. The regions ahead of the crack tips were systematically analysed as they have a potential to significantly influence crack propagation behaviour. This study shows a significant effect of carburisation at the crack tip compared to the bulk of the material, but only for pressurised CO₂ conditions, whereas atmospheric pressure CO₂ conditions more closely resembled the specimen exposed in air. Importantly, this study reveals that in simulated experiments, pressurised CO₂ environments are essential to replicate the carburisation behaviour observed in ex-service conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155625"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155249","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Thermo-migration of hydrogen isotopes in Eurofer and AISI 316L steel","authors":"Marta Malo ,&nbsp;Carlos Moreno ,&nbsp;Alejandro Moroño ,&nbsp;Igor Peñalva","doi":"10.1016/j.jnucmat.2025.155629","DOIUrl":"10.1016/j.jnucmat.2025.155629","url":null,"abstract":"<div><div>Permeation driven by thermal gradients (the Soret effect) was evaluated for H isotopes fusion grade steels, specifically Eurofer 97–3 and AISI 316 L as a separate process from the most commonly known permeation related to a gas concentration gradient. A unique permeation measurement system was developed to isolate and quantify permeation driven by thermal gradients, independent from the conventional concentration gradient-driven permeation. Experiments were conducted under fusion-relevant conditions, covering a pressure range of 0.1 mbar to 1000 mbar and temperatures between 300 °C to 500 °C, with thermal gradients up to 100 °C/mm.</div><div>The results show that permeation induced by thermal gradients can be significant, reaching magnitudes comparable to those of ordinary fluxes. These results indicate that the Soret effect should not be neglected in tritium retention and loss calculations, and suggest that future fusion reactor designs must incorporate these considerations for improved safety and efficiency. This research contributes a critical piece of knowledge to the field, with implications for both the development of fusion technology and the management of tritium in such systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155629"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155389","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}