Pub Date : 2024-10-24DOI: 10.1016/j.jnucmat.2024.155481
Aijia Lei , Xun Dai , Yufeng Du , Jingjing Liao , Ruiju Deng , Jiangtao Xu , Xuefei Huang
This study compared the uniform corrosion behavior of Zr-Sn-Fe-Cr-Ni alloys in deionized water at 360 °C/18.6 MPa with high and low concentration of dissolved oxygen (DO). It was found that high concentration of DO accelerated the corrosion rate of the Zr-Sn-Fe-Cr-Ni alloys and led to an earlier corrosion transition. Increased DO concentration resulted in a higher content of t-ZrO2 near the metal-oxide interface, which induced greater in-plane compressive stress in the oxide film and a high-level phase transformation from t-ZrO2 to m-ZrO2. This, in turn, led to an earlier occurrence of corrosion transition.
{"title":"Comparative studies of the long-term corrosion behavior of Zr-Sn-Fe-Cr-Ni alloys in pure water at 360 °C/18.6 MPa with high and low dissolved oxygen content","authors":"Aijia Lei , Xun Dai , Yufeng Du , Jingjing Liao , Ruiju Deng , Jiangtao Xu , Xuefei Huang","doi":"10.1016/j.jnucmat.2024.155481","DOIUrl":"10.1016/j.jnucmat.2024.155481","url":null,"abstract":"<div><div>This study compared the uniform corrosion behavior of Zr-Sn-Fe-Cr-Ni alloys in deionized water at 360 °C/18.6 MPa with high and low concentration of dissolved oxygen (DO). It was found that high concentration of DO accelerated the corrosion rate of the Zr-Sn-Fe-Cr-Ni alloys and led to an earlier corrosion transition. Increased DO concentration resulted in a higher content of t-ZrO<sub>2</sub> near the metal-oxide interface, which induced greater in-plane compressive stress in the oxide film and a high-level phase transformation from t-ZrO<sub>2</sub> to m-ZrO<sub>2</sub>. This, in turn, led to an earlier occurrence of corrosion transition.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155481"},"PeriodicalIF":2.8,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142554389","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}
Pub Date : 2024-10-24DOI: 10.1016/j.jnucmat.2024.155482
Ding Chen , Wei Dai , Daoxuan Liang , Qimin Wang , Jun Yan
Applying protective coatings to Zr alloy cladding surfaces is one of the better methods to design fuel tolerant materials. In this study, the surface of a Zr-4 alloy was coated with Cr using high-power impulse magnetron sputtering. Furthermore, the mechanisms by which bias voltages affect the mechanical characteristics, resistance to high-temperature steam oxidation, and coating structure were elucidated. The coating exhibits a strong (200) weave structure with coarse grains at a bias voltage of -100 V. With increasing bias, the energy of deposited particles increases, grains continue to grow, (200) preferential growth orientation disappears, and the coating exhibits a (110) crystal orientation. The growth structure of the coating first shows a tendency to be dense and then loose. For the Cr coating with a (200) crystal orientation, a dense oxide layer is preferentially formed after oxidation, which can effectively block the internal diffusion of O. With increasing oxidation time, coarse Cr grains can effectively block the external diffusion of Zr. Furthermore, the Cr coating exhibiting a (110) crystal orientation was severely oxidized after oxidation, resulting in the formation of cracks at the film base; this accelerated the outward diffusion of Zr.
{"title":"Mechanical and high-temperature steam oxidation properties of Cr coatings deposited via high-power impulse magnetron sputtering","authors":"Ding Chen , Wei Dai , Daoxuan Liang , Qimin Wang , Jun Yan","doi":"10.1016/j.jnucmat.2024.155482","DOIUrl":"10.1016/j.jnucmat.2024.155482","url":null,"abstract":"<div><div>Applying protective coatings to Zr alloy cladding surfaces is one of the better methods to design fuel tolerant materials. In this study, the surface of a Zr-4 alloy was coated with Cr using high-power impulse magnetron sputtering. Furthermore, the mechanisms by which bias voltages affect the mechanical characteristics, resistance to high-temperature steam oxidation, and coating structure were elucidated. The coating exhibits a strong (200) weave structure with coarse grains at a bias voltage of -100 V. With increasing bias, the energy of deposited particles increases, grains continue to grow, (200) preferential growth orientation disappears, and the coating exhibits a (110) crystal orientation. The growth structure of the coating first shows a tendency to be dense and then loose. For the Cr coating with a (200) crystal orientation, a dense oxide layer is preferentially formed after oxidation, which can effectively block the internal diffusion of O. With increasing oxidation time, coarse Cr grains can effectively block the external diffusion of Zr. Furthermore, the Cr coating exhibiting a (110) crystal orientation was severely oxidized after oxidation, resulting in the formation of cracks at the film base; this accelerated the outward diffusion of Zr.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155482"},"PeriodicalIF":2.8,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537329","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}
Pub Date : 2024-10-23DOI: 10.1016/j.jnucmat.2024.155471
J. Suárez-Recio , D. Fernández-Pello , M.A. Cerdeira , C. González , R. Gonzalez-Arrabal , R. Iglesias
Light impurity atoms (LIAs), such as hydrogen and helium, tend to aggregate at pre-existing intrinsic point defects. This aggregation leads to detrimental effects, particularly in environments such as those foreseen in nuclear fusion reactors. There, such impurities would be ubiquitous, resulting in unacceptable material behavior that would unqualify the material as a Plasma Facing Material (PFM). One option to delay the degradation in performance is the use of nanostructured tungsten (NW), showing a large density of grain boundaries (GBs). Although we have already addressed the behavior of a single LIA in a GB, in this work we present the combined synergistic effects of the simultaneous presence of multiple LIAs, vacancies and Self-Interstitial Atoms (SIA) at semicoherent W/W interfaces using ab initio methods. Our results reveal a complex and interesting process in the competition between LIAs and SIAs. When the number of SIAs is low, He appears to hinder their recombination with vacancies, therefore casting doubts on the self-healing provided by NW. However, in the presence of larger numbers of SIAs, their mutual repulsion leads to the opposite behavior. Thus, a thorough thermodynamic assessment in which the evolution of the system may be tracked emerges as the crucial subsequent step in these investigations.
轻杂质原子（LIAs），如氢和氦，往往会聚集在预先存在的固有点缺陷处。这种聚集会导致有害影响，尤其是在核聚变反应堆等环境中。在核聚变反应堆中，这种杂质会无处不在，导致不可接受的材料行为，从而使材料失去作为等离子体面层材料（PFM）的资格。延缓性能退化的一种方法是使用纳米结构的钨 (NW)，这种材料具有高密度的晶界 (GB)。虽然我们已经研究了 GB 中单个 LIA 的行为，但在这项工作中，我们使用 ab initio 方法，介绍了在半相干 W/W 界面同时存在多个 LIA、空位和自间隙原子 (SIA) 的综合协同效应。我们的研究结果揭示了 LIA 与 SIA 之间复杂而有趣的竞争过程。当 SIA 的数量较少时，He 似乎会阻碍它们与空位的重组，从而使人们对 NW 的自愈能力产生怀疑。然而，当 SIA 的数量较多时，它们之间的相互排斥会导致相反的行为。因此，进行彻底的热力学评估以跟踪系统的演变是这些研究的关键后续步骤。
{"title":"DFT simulations of the self-healing behavior of a W〈110〉/W〈112〉 grain boundary in the presence of coexisting point defects","authors":"J. Suárez-Recio , D. Fernández-Pello , M.A. Cerdeira , C. González , R. Gonzalez-Arrabal , R. Iglesias","doi":"10.1016/j.jnucmat.2024.155471","DOIUrl":"10.1016/j.jnucmat.2024.155471","url":null,"abstract":"<div><div>Light impurity atoms (LIAs), such as hydrogen and helium, tend to aggregate at pre-existing intrinsic point defects. This aggregation leads to detrimental effects, particularly in environments such as those foreseen in nuclear fusion reactors. There, such impurities would be ubiquitous, resulting in unacceptable material behavior that would unqualify the material as a Plasma Facing Material (PFM). One option to delay the degradation in performance is the use of nanostructured tungsten (NW), showing a large density of grain boundaries (GBs). Although we have already addressed the behavior of a single LIA in a GB, in this work we present the combined synergistic effects of the simultaneous presence of multiple LIAs, vacancies and Self-Interstitial Atoms (SIA) at semicoherent W/W interfaces using ab initio methods. Our results reveal a complex and interesting process in the competition between LIAs and SIAs. When the number of SIAs is low, He appears to hinder their recombination with vacancies, therefore casting doubts on the self-healing provided by NW. However, in the presence of larger numbers of SIAs, their mutual repulsion leads to the opposite behavior. Thus, a thorough thermodynamic assessment in which the evolution of the system may be tracked emerges as the crucial subsequent step in these investigations.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155471"},"PeriodicalIF":2.8,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142560922","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}
Pub Date : 2024-10-23DOI: 10.1016/j.jnucmat.2024.155477
Claire L. Corkhill , Latham T. Haigh , Lewis R. Blackburn , Luke T. Townsend , Daniel J. Bailey , Lucy M. Mottram , Amber R. Mason , Max R. Cole , Thierry Gervais , Genevieve Kerboul
The safe and secure management of civil separated plutonium is a UK government and NDA priority. One potential solution to address this considers the manufacture of a modified version of mixed oxide (MOX) fuel, comprising PuO2 dispersed within a UO2 matrix and doped with a suitable neutron absorbing element to maintain criticality control. As an initial step to understand whether an industrially-relevant, proven MOX fuel fabrication process could offer a potential route to the production of a Pu-disposition matrix based on MOX, a series of Gd-doped UO2 pellets were prepared by Orano at the CDA workshop of the MELOX facility in France. Characterisation was performed to quantify the density, morphology (grain size and porosity), Gd distribution and Gd incorporation mechanism. It was found that the materials produced were highly reproducible and similar in density and morphology, irrespective of the variables investigated, and similar to unirradiated UOX and MOX fuel. Gd was distributed in a similar manner to the distribution of PuO2 in unirradiated MIMAS (MIcronisation of a MASter Blend) MOX fuel and evidence for the existence of a solid solution between Gd2O3 and UO2 was ascertained, which could be viewed as favourable from a GDF post-closure criticality control perspective. The source of the powder had the greatest effect on the final characteristics of the Pu-disposition MOX pellets, due to sintering reactivity; however, these differences were minor. These results are a promising step towards the full-scale manufacture of ceramics suitable for the immobilisation and disposition of separated PuO2 in a GDF, should policy dictate.
安全可靠地管理民用分离钚是英国政府和国家原子能机构的优先事项。解决这一问题的一个潜在方案是考虑制造一种改进型混合氧化物（MOX）燃料，包括分散在二氧化铀基体中的二氧化铀，并掺入适当的中子吸收元素以保持临界控制。作为了解工业上相关的、经过验证的 MOX 燃料制造工艺能否为基于 MOX 的钚分散基质的生产提供潜在途径的第一步，奥拉诺公司在法国 MELOX 设施的 CDA 车间制备了一系列掺钆的二氧化铀颗粒。对密度、形态（晶粒大小和孔隙率）、钆分布和钆掺入机制进行了定量表征。研究发现，所生产的材料具有很高的可重复性，密度和形态相似，与未经过辐照的 UOX 和 MOX 燃料相似，与所研究的变量无关。钆的分布方式与未经过辐照的 MIMAS（MIcronisation of a MASter Blend）MOX 燃料中二氧 化钚的分布方式相似，并确定了 Gd2O3 和二氧铀之间存在固溶体的证据，这从 GDF 关闭后临界控制的角度来看是有利的。由于烧结反应性的原因，粉末来源对钚沉积 MOX 粒子的最终特性影响最大；不过，这些差异很小。这些结果是在政策允许的情况下，向全面制造适合固定和处置全球乏燃料发展基金中分离的二氧化铀的陶瓷迈出的充满希望的一步。
{"title":"Demonstration of industrially-fabricated plutonium disposition MOX","authors":"Claire L. Corkhill , Latham T. Haigh , Lewis R. Blackburn , Luke T. Townsend , Daniel J. Bailey , Lucy M. Mottram , Amber R. Mason , Max R. Cole , Thierry Gervais , Genevieve Kerboul","doi":"10.1016/j.jnucmat.2024.155477","DOIUrl":"10.1016/j.jnucmat.2024.155477","url":null,"abstract":"<div><div>The safe and secure management of civil separated plutonium is a UK government and NDA priority. One potential solution to address this considers the manufacture of a modified version of mixed oxide (MOX) fuel, comprising PuO<sub>2</sub> dispersed within a UO<sub>2</sub> matrix and doped with a suitable neutron absorbing element to maintain criticality control. As an initial step to understand whether an industrially-relevant, proven MOX fuel fabrication process could offer a potential route to the production of a Pu-disposition matrix based on MOX, a series of Gd-doped UO<sub>2</sub> pellets were prepared by Orano at the CDA workshop of the MELOX facility in France. Characterisation was performed to quantify the density, morphology (grain size and porosity), Gd distribution and Gd incorporation mechanism. It was found that the materials produced were highly reproducible and similar in density and morphology, irrespective of the variables investigated, and similar to unirradiated UOX and MOX fuel. Gd was distributed in a similar manner to the distribution of PuO<sub>2</sub> in unirradiated MIMAS (MIcronisation of a MASter Blend) MOX fuel and evidence for the existence of a solid solution between Gd<sub>2</sub>O<sub>3</sub> and UO<sub>2</sub> was ascertained, which could be viewed as favourable from a GDF post-closure criticality control perspective. The source of the powder had the greatest effect on the final characteristics of the Pu-disposition MOX pellets, due to sintering reactivity; however, these differences were minor. These results are a promising step towards the full-scale manufacture of ceramics suitable for the immobilisation and disposition of separated PuO<sub>2</sub> in a GDF, should policy dictate.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155477"},"PeriodicalIF":2.8,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571505","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}
Pub Date : 2024-10-22DOI: 10.1016/j.jnucmat.2024.155475
Da Wang , Weiqian Zhuo , Sirui Liu , Changquan Xiao , Wenjian Zhu , Bihan Sun , Xianfeng Ma , Ganfeng Yuan , Yulin Sun
This study investigated the compatibility of surface-nanostructured 15–15Ti austenitic steel in 550 °C LBE with an oxygen concentration of 5 × 10−7 wt.% for various exposure durations (759, 1638, 2404, and 3012 h). The results demonstrate that the grain size was reduced from 33.50 μm to the nano-scale after shot-peening (SP), achieving 17.62, 15.44, and 14.25 nm under SP pressures of 0.06, 0.15 and 0.25 MPa, respectively. The untreated steel experienced severe oxidation and dissolution corrosion, whereas the surface-nanostructured steel exhibited only mild oxidation and was resistant to dissolution corrosion. The enhanced corrosion resistance of surface-nanostructured steel is attributed to the higher protectiveness of the Cr-rich spinel layer and the less defective Ni-rich layer beneath it. Recrystallization occurred exclusively in the Ni-rich region, while the deformed steel underwent recovery during exposure. The thickness of the recrystallization layer was 2.9 μm at 759 h, increased to 8 μm at 1638 h, and remained stable thereafter. The size of recrystallized grains in SP-samples processed under pressure of 0.06 MPa and 0.15 MPa was approximately 2.92 μm, whereas it was about 1.32 μm for 0.25 MPa processed sample.
{"title":"Microstructure and corrosion property evolution of a surface-nanostructured 15–15Ti austenitic steel during immersion in liquid LBE at 550 °C","authors":"Da Wang , Weiqian Zhuo , Sirui Liu , Changquan Xiao , Wenjian Zhu , Bihan Sun , Xianfeng Ma , Ganfeng Yuan , Yulin Sun","doi":"10.1016/j.jnucmat.2024.155475","DOIUrl":"10.1016/j.jnucmat.2024.155475","url":null,"abstract":"<div><div>This study investigated the compatibility of surface-nanostructured 15–15Ti austenitic steel in 550 °C LBE with an oxygen concentration of 5 × 10<sup>−7</sup> wt.% for various exposure durations (759, 1638, 2404, and 3012 h). The results demonstrate that the grain size was reduced from 33.50 μm to the nano-scale after shot-peening (SP), achieving 17.62, 15.44, and 14.25 nm under SP pressures of 0.06, 0.15 and 0.25 MPa, respectively. The untreated steel experienced severe oxidation and dissolution corrosion, whereas the surface-nanostructured steel exhibited only mild oxidation and was resistant to dissolution corrosion. The enhanced corrosion resistance of surface-nanostructured steel is attributed to the higher protectiveness of the Cr-rich spinel layer and the less defective Ni-rich layer beneath it. Recrystallization occurred exclusively in the Ni-rich region, while the deformed steel underwent recovery during exposure. The thickness of the recrystallization layer was 2.9 μm at 759 h, increased to 8 μm at 1638 h, and remained stable thereafter. The size of recrystallized grains in SP-samples processed under pressure of 0.06 MPa and 0.15 MPa was approximately 2.92 μm, whereas it was about 1.32 μm for 0.25 MPa processed sample.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155475"},"PeriodicalIF":2.8,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142553739","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}
Matrix graphite (MG), a key component of fuel elements for high-temperature gas-cooled reactors (HTRs), has a profound effect on the comprehensive performance and service safety of fuel elements. A3–3 MG was selected as the matrix material for the pebble fuel elements of the 10 MW experimental high-temperature gas-cooled reactor (HTR-10) and the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) in China. During the preparation process of A3–3 MG, the green MG pebble must undergo two-stage heat treatment, namely carbonization and purification, to obtain excellent comprehensive properties for safe service. However, the porosity of A3–3 MG and its change during heat treatment remains unclear. Herein, the pore structure evolution through three different stages of A3–3 MG - the green, carbonized and purified samples- were tested using the gas adsorption method, mercury intrusion porosimetry and X-ray computed tomography (X-CT). The green sample had the smallest pore diameter and a uniform pore size distribution. The pore structure of the carbonized sample was the most developed, with the most micropores, mesopores and macropores. The molecular-sized micropores were produced due to the pyrogenic decomposition of the resin binder. Purification led to a decrease in pore diameter, together with a slight increase in closed pores and a decrease in pore connectivity due to pore merging and conversion. Two- and three-dimensional (2D and 3D) pore structure models were established by X-CT scan. The variation in pore size and shape, different types of pores as well as the pore conversion during the heat treatment process of A3–3 MG were observed. In this work, the porosity evolution of A3–3 MG was studied in detail, and references and strategies were provided for optimizing the preparation process and performance of pebble fuel elements.
{"title":"Pore structure evolution of A3–3 matrix graphite during heat treatment","authors":"Xi Tong, Xiangwen Zhou, Kaihong Zhang, Huixun Gao, Shouchi Zhang, Bing Liu, Yaping Tang","doi":"10.1016/j.jnucmat.2024.155474","DOIUrl":"10.1016/j.jnucmat.2024.155474","url":null,"abstract":"<div><div>Matrix graphite (MG), a key component of fuel elements for high-temperature gas-cooled reactors (HTRs), has a profound effect on the comprehensive performance and service safety of fuel elements. A3–3 MG was selected as the matrix material for the pebble fuel elements of the 10 MW experimental high-temperature gas-cooled reactor (HTR-10) and the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) in China. During the preparation process of A3–3 MG, the green MG pebble must undergo two-stage heat treatment, namely carbonization and purification, to obtain excellent comprehensive properties for safe service. However, the porosity of A3–3 MG and its change during heat treatment remains unclear. Herein, the pore structure evolution through three different stages of A3–3 MG - the green, carbonized and purified samples- were tested using the gas adsorption method, mercury intrusion porosimetry and X-ray computed tomography (X-CT). The green sample had the smallest pore diameter and a uniform pore size distribution. The pore structure of the carbonized sample was the most developed, with the most micropores, mesopores and macropores. The molecular-sized micropores were produced due to the pyrogenic decomposition of the resin binder. Purification led to a decrease in pore diameter, together with a slight increase in closed pores and a decrease in pore connectivity due to pore merging and conversion. Two- and three-dimensional (2D and 3D) pore structure models were established by X-CT scan. The variation in pore size and shape, different types of pores as well as the pore conversion during the heat treatment process of A3–3 MG were observed. In this work, the porosity evolution of A3–3 MG was studied in detail, and references and strategies were provided for optimizing the preparation process and performance of pebble fuel elements.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155474"},"PeriodicalIF":2.8,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142579110","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}
Pub Date : 2024-10-22DOI: 10.1016/j.jnucmat.2024.155467
Aleksandr Tsybanev , Alessandro Marino , Jun Lim , Kristof Gladinez , Nele Moelans
{"title":"Corrigendum to “Experimental assessment of thermodynamic stability and nucleation of NiO in liquid lead-bismuth eutectic for MYRRHA” [Journal of Nuclear Materials 603 (2025) 155404]","authors":"Aleksandr Tsybanev , Alessandro Marino , Jun Lim , Kristof Gladinez , Nele Moelans","doi":"10.1016/j.jnucmat.2024.155467","DOIUrl":"10.1016/j.jnucmat.2024.155467","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155467"},"PeriodicalIF":2.8,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537327","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}
Pub Date : 2024-10-21DOI: 10.1016/j.jnucmat.2024.155464
Pierre-Clément A. Simon, Jia-Hong Ke, Chao Jiang, Larry K. Aagesen, Wen Jiang
Tristructural isotropic (TRISO) particles are under consideration for use in several proposed advanced nuclear reactor concepts. The silicon carbide (SiC) layer in TRISO acts as a barrier to prevent the release of the fission products. However, despite remarkable retention, silver (Ag) release has been observed from intact particles, which requires investigation since the Ag isotope (Ag) has a long half-life. Previous work focused on developing a multiscale, mechanistic model for Ag diffusion accounting for temperature and microstructure effect and has been successfully validated. In this work, we expand the previous model to account for irradiation-enhanced Ag diffusivity in SiC and improve its accuracy over a wider grain size and temperature ranges relevant for advanced reactor conditions. A temperature, grain size, and flux dependent diffusivity is therefore derived using the mesoscale code MARMOT and implemented in the fuel performance code BISON. The irradiation-enhanced Ag diffusivity in SiC is compared against experimental data and validated using BISON against Ag release measurements from the Advanced Gas Reactor Fuel Development and Qualification Program (AGR-1 and AGR-2). Herein, we quantify the impact of SiC grain size, irradiation, and temperature on Ag release. In agreement with previous studies, we find accounting for SiC grain size improves agreement between BISON predictions and experimental observations for most cases. We also find that accounting for irradiation improves agreement for cases where Ag release was underestimated, but the impact was less significant than accounting for microstructure.
{"title":"Multiscale, mechanistic modeling of irradiation-enhanced silver diffusion in TRISO particles","authors":"Pierre-Clément A. Simon, Jia-Hong Ke, Chao Jiang, Larry K. Aagesen, Wen Jiang","doi":"10.1016/j.jnucmat.2024.155464","DOIUrl":"10.1016/j.jnucmat.2024.155464","url":null,"abstract":"<div><div>Tristructural isotropic (TRISO) particles are under consideration for use in several proposed advanced nuclear reactor concepts. The silicon carbide (SiC) layer in TRISO acts as a barrier to prevent the release of the fission products. However, despite remarkable retention, silver (Ag) release has been observed from intact particles, which requires investigation since the Ag isotope (<span><math><msup><mrow></mrow><mrow><mn>110</mn><mi>m</mi></mrow></msup></math></span>Ag) has a long half-life. Previous work focused on developing a multiscale, mechanistic model for Ag diffusion accounting for temperature and microstructure effect and has been successfully validated. In this work, we expand the previous model to account for irradiation-enhanced Ag diffusivity in SiC and improve its accuracy over a wider grain size and temperature ranges relevant for advanced reactor conditions. A temperature, grain size, and flux dependent diffusivity is therefore derived using the mesoscale code MARMOT and implemented in the fuel performance code BISON. The irradiation-enhanced Ag diffusivity in SiC is compared against experimental data and validated using BISON against Ag release measurements from the Advanced Gas Reactor Fuel Development and Qualification Program (AGR-1 and AGR-2). Herein, we quantify the impact of SiC grain size, irradiation, and temperature on Ag release. In agreement with previous studies, we find accounting for SiC grain size improves agreement between BISON predictions and experimental observations for most cases. We also find that accounting for irradiation improves agreement for cases where Ag release was underestimated, but the impact was less significant than accounting for microstructure.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155464"},"PeriodicalIF":2.8,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142553740","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}
Pub Date : 2024-10-21DOI: 10.1016/j.jnucmat.2024.155473
Dewang Cui , Shuo Cong , Ziqi Cao , Fan Yuan , Guang Ran
The combined effects of corrosion and irradiation on nuclear components have been an important but not yet fully revealed topic. Here, the irradiation behavior of the oxide scale formed on F/M steel after lead-bismuth corrosion was in-situ investigated during He+ irradiation. The results showed that the oxide scale included a Fe3O4 outer oxide layer, a nanograin Fe(FexCr2-x)O4 spinel inner oxide layer, and an internal oxide layer. He bubbles formed in Fe3O4, Fe-Cr spinel and F/M steel were polygon, irregular elongated pores and small spheres, respectively. These differences were attributed to variations in defect generation, migration, and corrosion-induced crystal defects in different oxides. Numerous corrosion-induced nanograin boundaries and vacancies in Fe-Cr spinel exhibited more effective absorption of irradiation-induced defects. Moreover, rhombic perfect dislocation loops were detected in Fe3O4 at the late stage of irradiation, their relative positional relationship with He bubbles indicated a potential interaction between bubbles and loops.
腐蚀和辐照对核部件的综合影响一直是一个重要但尚未完全揭示的课题。在此，我们对铅铋腐蚀后在 F/M 钢上形成的氧化鳞在 He+ 辐照下的辐照行为进行了原位研究。结果表明，氧化鳞包括外氧化层 Fe3O4、纳米晶粒 Fe(FexCr2-x)O4 尖晶石内氧化层和内部氧化层。在 Fe3O4、Fe-Cr 尖晶石和 F/M 钢中形成的 He 气泡分别为多边形、不规则细长孔隙和小球形。这些差异归因于不同氧化物中缺陷生成、迁移和腐蚀诱导晶体缺陷的变化。铁铬尖晶石中大量的腐蚀诱导纳米晶界和空位对辐照诱导缺陷的吸收更为有效。此外，在辐照后期，在 Fe3O4 中检测到菱形完美位错环，它们与 He 气泡的相对位置关系表明气泡与环之间可能存在相互作用。
{"title":"In-situ study on the differential evolution of He bubbles in the multilayer oxide of Fe9Cr1.5W0.4Si F/M steel corroded in lead-bismuth eutectic","authors":"Dewang Cui , Shuo Cong , Ziqi Cao , Fan Yuan , Guang Ran","doi":"10.1016/j.jnucmat.2024.155473","DOIUrl":"10.1016/j.jnucmat.2024.155473","url":null,"abstract":"<div><div>The combined effects of corrosion and irradiation on nuclear components have been an important but not yet fully revealed topic. Here, the irradiation behavior of the oxide scale formed on F/M steel after lead-bismuth corrosion was in-situ investigated during He<sup>+</sup> irradiation. The results showed that the oxide scale included a Fe<sub>3</sub>O<sub>4</sub> outer oxide layer, a nanograin Fe(Fe<sub>x</sub>Cr<sub>2-x</sub>)O<sub>4</sub> spinel inner oxide layer, and an internal oxide layer. He bubbles formed in Fe<sub>3</sub>O<sub>4</sub>, Fe-Cr spinel and F/M steel were polygon, irregular elongated pores and small spheres, respectively. These differences were attributed to variations in defect generation, migration, and corrosion-induced crystal defects in different oxides. Numerous corrosion-induced nanograin boundaries and vacancies in Fe-Cr spinel exhibited more effective absorption of irradiation-induced defects. Moreover, rhombic perfect dislocation loops were detected in Fe<sub>3</sub>O<sub>4</sub> at the late stage of irradiation, their relative positional relationship with He bubbles indicated a potential interaction between bubbles and loops.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155473"},"PeriodicalIF":2.8,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142554390","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}
Pub Date : 2024-10-21DOI: 10.1016/j.jnucmat.2024.155461
Yaojun Li , Sirui Liu , Hailin Zhai , Yuexia Wang , Yan Zhao , Xianfeng Ma
The first-principles were employed to investigate the structure, adhesion, and tensile properties of the Zr(0001)/SiC close-packed interface with different vacancies. From the perspective of vacancy formation energy, SiC coating is beneficial for enhancing the irradiation resistance of Zr cladding. When vacancies are present, except for the Zr2 and Zr3 vacancies, introducing other vacancies reduces the stability of Zr/SiC interfaces. The C-terminated interface is more stable than the Si-terminated interface. Through electronic structure analysis, vacancies at the interface primarily reduce the bonds between Zr and SiC, decreasing the interface stability. Vacancies on the side of SiC indirectly alter the strength or quantity of covalent (bonding or anti-bonding) and ionic bonds at the interface, thus intricately lowering the interface stability. In tensile tests, the cleavage of all interfaces with vacancies still occurs on the side of Zr. Vacancies on the SiC side partly lead to increased electrons between Zr1-Zr2 or Zr2-Zr3, strengthening the metallic bonds and enhancing the interface's ideal strength and ductility. The present study offers a novel perspective from the standpoint of bonding mechanisms, providing good insights into the effects of different vacancies on the performance of Zr/SiC interfaces.
{"title":"The influence of different vacancies on Zr(0001)/SiC close-packed interface performance: A first-principles study","authors":"Yaojun Li , Sirui Liu , Hailin Zhai , Yuexia Wang , Yan Zhao , Xianfeng Ma","doi":"10.1016/j.jnucmat.2024.155461","DOIUrl":"10.1016/j.jnucmat.2024.155461","url":null,"abstract":"<div><div>The first-principles were employed to investigate the structure, adhesion, and tensile properties of the Zr(0001)/SiC close-packed interface with different vacancies. From the perspective of vacancy formation energy, SiC coating is beneficial for enhancing the irradiation resistance of Zr cladding. When vacancies are present, except for the Zr2 and Zr3 vacancies, introducing other vacancies reduces the stability of Zr/SiC interfaces. The C-terminated interface is more stable than the Si-terminated interface. Through electronic structure analysis, vacancies at the interface primarily reduce the bonds between Zr and SiC, decreasing the interface stability. Vacancies on the side of SiC indirectly alter the strength or quantity of covalent (bonding or anti-bonding) and ionic bonds at the interface, thus intricately lowering the interface stability. In tensile tests, the cleavage of all interfaces with vacancies still occurs on the side of Zr. Vacancies on the SiC side partly lead to increased electrons between Zr1-Zr2 or Zr2-Zr3, strengthening the metallic bonds and enhancing the interface's ideal strength and ductility. The present study offers a novel perspective from the standpoint of bonding mechanisms, providing good insights into the effects of different vacancies on the performance of Zr/SiC interfaces.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155461"},"PeriodicalIF":2.8,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571504","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}
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