Journal of Nuclear Materials最新文献

{"title":"Effect of microstructural features on tensile properties of β-Quenched Zr-2.5 %Nb alloy","authors":"Aishwarya Upadhye ,&nbsp;Arnomitra Chatterjee ,&nbsp;A.K. Bind ,&nbsp;Saurav Sunil ,&nbsp;Apu Sarkar ,&nbsp;R.N. Singh","doi":"10.1016/j.jnucmat.2025.155958","DOIUrl":"10.1016/j.jnucmat.2025.155958","url":null,"abstract":"<div><div>Zr-2.5 %Nb alloy in the cold-worked and stress-relieved (CWSR) condition is used as pressure tube (PT) material in Pressurized Heavy Water Reactors (PHWRs) with a design life of 30 years. To enhance in-reactor performance, an alternative fabrication route involving heat treatment is being explored for Indian PHWRs<strong>.</strong> For the development of heat-treated pressure tubes, it is essential to optimize process parameters to achieve the desired microstructure and texture. Among microstructural factors, prior β-grain size is crucial in determining the mechanical properties of the material. In this study, Zr-2.5 %Nb alloy specimens were exposed to varying temperatures within the β-phase region for different soaking times. Rapid heating and quenching techniques were employed to ensure the formation of a random texture. The prior β-grain size was assessed using optical microscopy, while the martensitic lath structure was characterized through electron backscatter diffraction (EBSD). The tensile properties of the heat-treated specimens were evaluated at room temperature. The study tracked the evolution of prior β-grain size over time to identify different stages of grain growth. This work investigates the influence of prior β-grain size, martensitic lath structure, and dislocation density on the tensile properties of the β-quenched Zr-2.5 %Nb alloy.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155958"},"PeriodicalIF":2.8,"publicationDate":"2025-06-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144223235","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":"Impact of anisotropic grain microstructure on the fracture-mechanical properties of ITER grade tungsten plate","authors":"E. Gaganidze,&nbsp;N. Meena,&nbsp;L. Chauhan,&nbsp;Q. Yuan,&nbsp;J. Aktaa","doi":"10.1016/j.jnucmat.2025.155945","DOIUrl":"10.1016/j.jnucmat.2025.155945","url":null,"abstract":"<div><div>The fracture-mechanical behavior of a uniaxially rolled ITER grade tungsten plate has been investigated up to test temperatures of 700°C. To study the influence of the anisotropic grain microstructure on the Ductile-to-Brittle Transition Temperature (DBTT) and failure mechanism, fracture toughness testing has been performed for four specimen extraction orientations: longitudinal - long transverse (L-T), long transverse - longitudinal (T-L), longitudinal - short transverse (L-S) and long transverse - short transverse (T-S). The DBTTs were at around 250°C and 300°C in L-T and T-L orientations, respectively. In the case of L-S and T-S orientations, the crack deflection prevented unambiguous identification of the DBTT. An upper bound of the DBTT for L-S and T-S orientations was estimated in the range of 300-350°C. Transgranular cleavage was a dominant fracture mode for L-T and T-L orientations at low test temperatures. In contrast to this, for L-S and T-S orientations strong deflections of the crack towards the rolling direction have been recognized which are linked to the anisotropic grain microstructure of the plate material. At the highest test temperatures, no failure of the specimens has been observed up to severe deflection levels which is linked to an extended blunting of the crack starter notch tip and related suppression of the crack initiation and propagation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155945"},"PeriodicalIF":2.8,"publicationDate":"2025-06-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144279113","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":"Efficient adsorption of radioactive iodine using silver-incorporated bentonite–chitosan hydrogel: Batch and column studies","authors":"Byungjun Cha ,&nbsp;Yeonji Yea ,&nbsp;Lewis Kamande Njaramba ,&nbsp;Minseok Kim ,&nbsp;Yeomin Yoon ,&nbsp;Chang Min Park","doi":"10.1016/j.jnucmat.2025.155943","DOIUrl":"10.1016/j.jnucmat.2025.155943","url":null,"abstract":"<div><div>The discharge of radioactive iodine, particularly through nuclear wastewater, poses a persistent environmental threat, which demands effective remediation strategies. In this study, we investigated a novel chitosan-bentonite-silver (CBA) hydrogel designed to efficiently remove radioactive iodine. The bentonite-silver (Ben-Ag) composite was synthesized using a straightforward ion-exchange method and subsequently embedded in a chitosan matrix to form the hydrogel structure. The synthesized CBA hydrogels were systematically characterized using multiple analytical techniques and evaluated for iodide removal performance through batch and continuous column studies. The optimized hydrogel, CBA50, exhibited a notable maximum adsorption capacity of 21.6 mg/g, corresponding best to the Langmuir isotherm model, and kinetic analysis revealed that the data aligned closely with the PSO kinetic model, which suggests a chemisorption-driven adsorption process. Moreover, the hydrogels showed robust adsorption performance across a broad pH spectrum (pH 4–11) and in various anions and cations coexisting environments. Regeneration experiments exhibited considerable removal efficiency over five successive cycles, which highlights the potential reuse of hydrogels. Several removal mechanisms involving precipitation, electrostatic interactions, ion exchange, pore filling, and hydrogen bonding were proposed. These findings suggest that the CBA hydrogels are both efficient and environmentally friendly adsorbents for the removal of radioactive iodine in wastewater.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155943"},"PeriodicalIF":2.8,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144230125","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Evaluation of AGR-3/4 In-pile Silver Release Predictions Against Post-Irradiation Examination Measurements","authors":"William F. Skerjanc ,&nbsp;Wen Jiang ,&nbsp;Paul A. Demkowicz ,&nbsp;John D. Stempien","doi":"10.1016/j.jnucmat.2025.155942","DOIUrl":"10.1016/j.jnucmat.2025.155942","url":null,"abstract":"<div><div>Fuel performance modeling codes that accurately predict the transport of radionuclides in high-temperature gas-cooled reactors that utilize tristructural isotopic (TRISO) fuel particles are an important aspect of reactor safety analyses. One objective of the Advanced Gas Reactor (AGR)-3/4 experiment was to assess the transport of fission products through fuel particles and their subsequent release into the compact matrix and structural graphite materials. This was accomplished by irradiating uranium oxycarbide (UCO) driver fuel particles and designed-to-fail (DTF) particles to serve as known sources of fission products. The fission product of particular interest when it comes to such transport is silver (Ag-110 m), as it has a 250-day half-life and has relatively high mobility in the TRISO coating layers. To assess the current modeling capabilities and diffusion parameters employed in the fuel performance codes PARFUME and BISON, the fractional release of silver release predicted by the two codes were compared against post-irradiation examination measurements from the AGR-3/4 experiment.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155942"},"PeriodicalIF":2.8,"publicationDate":"2025-05-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144212971","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":"Oxidation resistance of high-burnup Cr-doped UO2 accident tolerant fuel and comparison with irradiated UO2","authors":"A. Milena-Pérez ,&nbsp;J.M. Elorrieta ,&nbsp;L. Emblico ,&nbsp;L.J. Bonales ,&nbsp;D. Serrano-Purroy ,&nbsp;N. Rodríguez-Villagra ,&nbsp;H. Galán","doi":"10.1016/j.jnucmat.2025.155930","DOIUrl":"10.1016/j.jnucmat.2025.155930","url":null,"abstract":"<div><div>The oxidation behaviour of a high-burnup Cr-doped UO₂ fuel under dry interim storage conditions has been <em>in-situ</em> analysed by Raman spectroscopy. In particular, this shortest-term accident tolerant fuel (ATF) has been subjected to 350 ºC in air for about 1000 h and its oxidation reaction compared to that of a UO₂ spent fuel irradiated in the same conditions and exposed to the same thermal treatment. The results obtained here do not only sustain the previous observation of a delay in U₃O₈ formation in the outer periphery (rim area), compared to the pellet centre for high-burnup spent fuels, but they also prove for the first time the greater resistance of this irradiated ATF to the complete oxidation of its UO₂ matrix. These findings have been confirmed with a further test of the Cr-doped UO₂ fuel at 400 ºC, where its significant heterogeneity has been additionally noticed by the continuous acquisition of Raman spectra in different regions of the pellet.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155930"},"PeriodicalIF":2.8,"publicationDate":"2025-05-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144202947","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":"Stress-induced corrosion behavior of martensitic P91 steel in high-temperature and high-pressure supercritical carbon dioxide for brayton cycle system","authors":"Shengxu Wang,&nbsp;Yingying Yang,&nbsp;Qiguo Yang,&nbsp;Aizheng Li,&nbsp;Yan Ren,&nbsp;Weidong Wu","doi":"10.1016/j.jnucmat.2025.155939","DOIUrl":"10.1016/j.jnucmat.2025.155939","url":null,"abstract":"<div><div>The corrosion behavior of alloy materials employed in high-temperature components of supercritical carbon dioxide (S-CO₂) Brayton cycle power generation systems is a critical factor influencing both the efficiency and service life of the system. In this study, we investigated the corrosion behavior of martensitic P91 steel exposed to S-CO₂ at 550 °C and 20 MPa for up to 1000 h. The effect of stress loading (0/210/420 MPa) on the corrosion behavior was studied using a four-point bending stress loading device. The results revealed that the surface of the specimen exhibited extensive corrosion areas, indicating that the corrosion type of martensitic P91 was uniform corrosion rather than localized corrosion. In addition, the corrosion layer on the surface displayed a carbon deposition phenomenon, which differs from the metal corrosion typically observed in conventional steam cycles. The thickness of the corrosion product layer on the specimen surface under three conditions were 15.6 μm, 17.6 μm, and 27 μm, respectively. All corrosion layers exhibited a double-layer structure, with the outer layer primarily composed of Fe₃O₄ and the inner layer mainly consisting of FeCr₂O₄ and Fe₂SiO₄. This indicates that stress loading has minimal impact on the phase composition of the corrosion products, but it accelerates the corrosion rate of the steel, increasing the thickness of the oxide layer. Additionally, the oxide layer thickness increases with rising stress values, accompanied by localized spalling.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155939"},"PeriodicalIF":2.8,"publicationDate":"2025-05-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144212969","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":"Proton irradiation-induced microstructure changes in a CrFeMnNi high entropy alloy","authors":"Qiang Wang ,&nbsp;Hui Yuan ,&nbsp;Cong Dai ,&nbsp;Hongbing Yu ,&nbsp;Amy Fluke ,&nbsp;Travis Skippon ,&nbsp;Fei Long ,&nbsp;Suraj Y. Persaud ,&nbsp;Mark R. Daymond ,&nbsp;Yanwen Zhang","doi":"10.1016/j.jnucmat.2025.155940","DOIUrl":"10.1016/j.jnucmat.2025.155940","url":null,"abstract":"<div><div>High entropy alloys have demonstrated superior radiation tolerance compared to pure metals and some conventional alloys. However, the stability of high entropy solid solution microstructure within a cascade-involved irradiation environment at elevated temperatures has not been well understood. In this study, we examined the microstructural evolution of a CrFeMnNi high entropy alloy (HEA) subjected to irradiation with 2 MeV protons under three different conditions: 2.8 dpa at 400 °C, 2.8 dpa at 600 °C, and 16.8 dpa at 400 °C. The irradiation-induced microstructural changes were characterized using (scanning) transmission electron microscopy ((S)TEM) combined with energy dispersive x-ray spectrometry (EDS) and electron energy loss spectroscopy (EELS). Irradiation-induced Frank loops in the damage plateau areas were statistically characterized. The samples irradiated to 2.8 dpa at 400 °C exhibited the highest density of Frank loops, whereas the largest average size of Frank loops was observed in the sample irradiated at 600 °C. Spinodal decomposition and L1₀-type NiMn ordering were observed in the irradiated samples, with the extent of decomposition increasing with irradiation temperature. Additionally, segregation of Ni and Fe and depletion of Mn were observed around Frank loops. Voids were observed only within the Cr precipitates in the 600 °C sample, which is attributed to the excess vacancies associated with Cr precipitation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155940"},"PeriodicalIF":2.8,"publicationDate":"2025-05-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144261557","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":"Effects of single and dual beam Fe and He ion irradiation on NbMoTaW and Ti2ZrHfV0.5Mo0.2 refractory high-entropy alloys","authors":"Peng Zhang ,&nbsp;Na Li ,&nbsp;Xiaonan Zhang ,&nbsp;Jinhua Hao ,&nbsp;Xianxiu Mei","doi":"10.1016/j.jnucmat.2025.155936","DOIUrl":"10.1016/j.jnucmat.2025.155936","url":null,"abstract":"<div><div>This study compared the behavior of two refractory high-entropy alloys (RHEAs), NbMoTaW and Ti₂ZrHfV_0.5Mo_0.2 (Nb RHEA and Ti RHEA), under single-beam Fe ion, single-beam He ion, and dual-beam irradiation, with a focus on He bubble evolution, dislocation generation, and irradiation hardening. Microscopic characterization revealed that, under both single-beam He ion and dual-beam irradiation, the size of He bubbles in Ti RHEA was larger than in Nb RHEA. Under dual-beam irradiation, the high displacement per atom (DPA) induced by Fe ion implantation promoted the growth of He bubbles, resulting in larger He bubble sizes in both RHEAs compared to those observed under single-beam He ion irradiation. The two RHEAs exhibited distinct dislocation evolution behaviors. Under single-beam ion irradiation, small, dense dislocation loops formed in the low DPA region, while dislocation line networks were observed in the high DPA region of Nb RHEA. Conversely, in Ti RHEA, high-density small dislocation loops were observed only in the damage peak region, while no evident irradiation-induced dislocation loops were observed in the front or attenuation regions. Under dual-beam irradiation, He clusters/He bubbles were observed to pin dislocations in both RHEAs. Nanoindentation results showed that the hardening of Nb RHEA was greater than in Ti RHEA. Nb RHEA exhibited the greatest hardening under single-beam He ion irradiation, while Ti RHEA exhibited the greatest hardening under dual-beam irradiation, which correlated with the distinct dislocation evolution behavior in the two alloys.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155936"},"PeriodicalIF":2.8,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144212970","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":"Evolution of irradiated Pile Grade A graphite microstructure under novel molten salt decontamination conditions","authors":"Tatiana Grebennikova ,&nbsp;Iuliia Ipatova ,&nbsp;Daniel N.T. Barton ,&nbsp;Robert N. Worth ,&nbsp;Ben F. Spencer ,&nbsp;Clint A. Sharrad ,&nbsp;Abbie N. Jones","doi":"10.1016/j.jnucmat.2025.155935","DOIUrl":"10.1016/j.jnucmat.2025.155935","url":null,"abstract":"<div><div>The irradiated graphite waste stream represents a significant challenge for nuclear power plant decommissioning in the UK, with an estimated 96,000 tonnes of graphite waste arising from the shutdown of the UK's gas-cooled reactors. The removal of activated impurities and fission products from irradiated graphite has been successfully performed previously [1] using an electrochemical molten salt decontamination approach. In this study, the material behaviour and structural changes of the treated nuclear graphite under molten salt treatment conditions have been assessed using multi-technique characterisation (Brunauer–Emmett–Teller surface area, Scanning Electron Microscopy, X-ray Photoelectron Spectroscopy and X-ray Diffraction). This novel research highlights that molten salt treatment leads to changes to the binder and impregnated phases while leaving the filler particles intact under the researched treatment conditions. Significant differences in atomic concentrations of C 1s deconvoluted peaks were observed, suggesting that the mechanism involved both diffusion of pre-adsorbed oxygen and limited chlorination of the surface during the decontamination process. The stability of the lattice parameters and minimal change in crystalline dimensions in molten salt-treated graphite material combined with limited mass loss provides a first-of-a-kind insight into the mechanisms behind graphite decontamination using the electrochemical molten salt approach. The findings support the future potential for wide-scale irradiated graphite treatment to achieve waste volume reduction, minimisation and re-categorisation in line with the waste hierarchy.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155935"},"PeriodicalIF":2.8,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144240953","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":"Scalable fabrication and deuterium release properties of Li2TiO3-Li2ZrO3 ceramic pebbles via centrifugal granulation","authors":"Guangfan Tan ,&nbsp;Qilai Zhou ,&nbsp;Yasuhisa Oya ,&nbsp;Yingchun Zhang ,&nbsp;Chang-An Wang ,&nbsp;Yanhao Dong","doi":"10.1016/j.jnucmat.2025.155933","DOIUrl":"10.1016/j.jnucmat.2025.155933","url":null,"abstract":"<div><div>The Li₂TiO₃ and Li₂ZrO₃ ceramic pebbles are regarded as promising tritium breeder on account of their superb chemical stability, high tritium diffusion rates, and neutron multiplication potential of Zr, respectively. To integrate the advantages of these two materials, it is necessary to develop a composite such as Li₂TiO₃-Li₂ZrO₃, which is ideal for advanced fusion reactor designs. Up to now, fabricated biphasic ceramic pebbles generally suffer from low production yield, inadequate crushing load, and high cost, which restricts their application in practice. In this paper, biphasic Li₂TiO₃-Li₂ZrO₃ tritium breeding pebbles with superior crushing strength are successfully fabricated employing the centrifugal granulation method by selecting the appropriate binder and controlling the growth rate. Moreover, by this technique, ceramic pebbles with wide diameter distribution can be obtained, which will help to improve the packing factor of pebble beds in the solid blanket. In addition, the composition, microstructure, and internal structure of the Li₂TiO₃-Li₂ZrO₃ ceramic pebbles are thoroughly analyzed, respectively. The Li₂TiO₃-Li₂ZrO₃ ceramic pebbles, after being sintered at 1100 °C, attained a high sphericity of 0.97, a superior crushing load of 67.6 N, and an optimum porosity of 9.75 % in the shell. Moreover, the Li₂TiO₃-Li₂ZrO₃ ceramic pebbles also exhibit excellent deuterium release properties with the main form of HDO. The above results show that the centrifugal granulation method is not only applicable to the mass production of Li₂TiO₃-Li₂ZrO₃ ceramic pebbles but also to other tritium breeders, which offers promising prospects for the development of advanced tritium breeding materials in the future.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 155933"},"PeriodicalIF":2.8,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144202946","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}