Pub Date : 2024-10-20DOI: 10.1016/j.jnucmat.2024.155472
Hui Dan, Yihang Li, Bingbing Bao, Jiajing Li, Jiyuan Guo, Yi Ding
Developing new matrix for efficient actinides immobilization is of great significance for the sustainable development of nuclear energy. Herein, novel ZrSiO4-borosilicate glass-ceramics (Z-B) were prepared for immobilization of cerium (Ce) as the simulated tetravalent actinides. The effect of Ce content on the phase transformation and microstructure of the obtained Z-B was investigated, and the loading capacity limit of Ce was evaluated. The results demonstrated that Z-B glass-ceramics with high ZrSiO4 phase (91 wt%) was obtained. Owing to the synergistic effect of ZrSiO4 and borosilicate glass phases, the loading capacity limit of Ce in the obtained Z-B reached up to 12 at%. Furthermore, the obtained Z-B waste forms exhibited excellent aqueous durability. The results of this work demonstrated that the Z-B is potential matrix for immobilization of tetravalent actinides due to their good loading capacity and aqueous durability.
{"title":"Development of ZrSiO4-borosilicate glass-ceramics for immobilization of simulated tetravalent actinides","authors":"Hui Dan, Yihang Li, Bingbing Bao, Jiajing Li, Jiyuan Guo, Yi Ding","doi":"10.1016/j.jnucmat.2024.155472","DOIUrl":"10.1016/j.jnucmat.2024.155472","url":null,"abstract":"<div><div>Developing new matrix for efficient actinides immobilization is of great significance for the sustainable development of nuclear energy. Herein, novel ZrSiO<sub>4</sub>-borosilicate glass-ceramics (Z-B) were prepared for immobilization of cerium (Ce) as the simulated tetravalent actinides. The effect of Ce content on the phase transformation and microstructure of the obtained Z-B was investigated, and the loading capacity limit of Ce was evaluated. The results demonstrated that Z-B glass-ceramics with high ZrSiO<sub>4</sub> phase (91 wt%) was obtained. Owing to the synergistic effect of ZrSiO<sub>4</sub> and borosilicate glass phases, the loading capacity limit of Ce in the obtained Z-B reached up to 12 at%. Furthermore, the obtained Z-B waste forms exhibited excellent aqueous durability. The results of this work demonstrated that the Z-B is potential matrix for immobilization of tetravalent actinides due to their good loading capacity and aqueous durability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155472"},"PeriodicalIF":2.8,"publicationDate":"2024-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537330","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 : 2024-10-18DOI: 10.1016/j.jnucmat.2024.155452
M.W.D. Cooper, C. Matthews, D.A. Andersson
Fission gas bubbles in UO2 nuclear fuel have been observed to exhibit pressures in excess of the equilibrium bubble pressure; however, the cause of bubble over-pressurization has not yet been demonstrated. The mechanical interaction between a bubble and the surrounding matrix or grain boundary depends on the internal pressure of the bubble and local stress state, such that over-pressurized bubbles are thought to be responsible for fragmentation and pulverization, when exposed to a temperature ramp. Here, we investigate the role of U interstitials, produced through irradiation, in over-pressurizing bubbles by using a combined molecular dynamics (MD) and cluster dynamics approach. Firstly, the energies for the capture of interstitials and vacancies by bubbles have been determined from MD as a function of the ratio of gas atoms to vacancies that make up the bubble. Secondly, these reaction energies have been implemented in the cluster dynamics code Centipede to predict bubble over-pressurization as a function of temperature for typical fission rates. It was found that there is a transition from low pressure bubbles (at high temperatures) to high pressure bubbles (at lower temperatures). The cause of this behavior was shown to be the creation of irradiation-induced interstitials that are highly mobile relative to vacancies at low temperature; whereas, vacancies are sufficiently mobile at high temperatures to limit bubble pressures. This result supports the hypothesis that over-pressurized bubbles form during steady-state operation and that this behavior is highly sensitive to the local pellet temperature.
{"title":"The role of irradiation-enhanced interstitial diffusion in over-pressurizing fission gas bubbles in UO2","authors":"M.W.D. Cooper, C. Matthews, D.A. Andersson","doi":"10.1016/j.jnucmat.2024.155452","DOIUrl":"10.1016/j.jnucmat.2024.155452","url":null,"abstract":"<div><div>Fission gas bubbles in UO<sub>2</sub> nuclear fuel have been observed to exhibit pressures in excess of the equilibrium bubble pressure; however, the cause of bubble over-pressurization has not yet been demonstrated. The mechanical interaction between a bubble and the surrounding matrix or grain boundary depends on the internal pressure of the bubble and local stress state, such that over-pressurized bubbles are thought to be responsible for fragmentation and pulverization, when exposed to a temperature ramp. Here, we investigate the role of U interstitials, produced through irradiation, in over-pressurizing bubbles by using a combined molecular dynamics (MD) and cluster dynamics approach. Firstly, the energies for the capture of interstitials and vacancies by bubbles have been determined from MD as a function of the ratio of gas atoms to vacancies that make up the bubble. Secondly, these reaction energies have been implemented in the cluster dynamics code Centipede to predict bubble over-pressurization as a function of temperature for typical fission rates. It was found that there is a transition from low pressure bubbles (at high temperatures) to high pressure bubbles (at lower temperatures). The cause of this behavior was shown to be the creation of irradiation-induced interstitials that are highly mobile relative to vacancies at low temperature; whereas, vacancies are sufficiently mobile at high temperatures to limit bubble pressures. This result supports the hypothesis that over-pressurized bubbles form during steady-state operation and that this behavior is highly sensitive to the local pellet temperature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155452"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652837","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 : 2024-10-18DOI: 10.1016/j.jnucmat.2024.155465
Jun Zhou, Nancy Lai Mun Wong, Jianwei Chai, Shijie Wang
B4C is an important material in diverse nuclear applications. However, a systematic examination of its surface properties is still missing. In this work, we employ first-principles simulations to investigate the energetic stability of 16 distinct slab models representing (001), (100), (101), (110), and (111) surfaces, which are constructed by minimizing dangling bonds. Our results show that C-terminated (001) surface exhibits significantly greater stability than other surfaces under both the carbon and boron-rich conditions. Besides, we also study the defect formation energies on the C-terminated (001) surface and compare them with the cases in bulk. The high formation energies of the defects suggest a low likelihood of their occurrence on this surface, despite their formation energies being lower compared to bulk cases. Furthermore, mid-gap surface states are revealed for the top atomic layers of the C-terminated (001) surface, which are deduced at the deeper layers, and the band structures of the middle layers of this slab recover to the bulk band gap. These surface mid-gap states allow electron excitation from the valence band to these states, resulting in a reduced optical band gap compared to the bulk band gap of B4C. This provides a plausible explanation for the significantly smaller band gap observed in experiments compared to the larger gap predicted by theoretical models. Our study not only sheds light on the surface properties of B4C but also lays the groundwork for advancing this material for more advanced nuclear applications.
{"title":"An investigation on the surface properties of B4C for advancing its nuclear applications","authors":"Jun Zhou, Nancy Lai Mun Wong, Jianwei Chai, Shijie Wang","doi":"10.1016/j.jnucmat.2024.155465","DOIUrl":"10.1016/j.jnucmat.2024.155465","url":null,"abstract":"<div><div>B<sub>4</sub>C is an important material in diverse nuclear applications. However, a systematic examination of its surface properties is still missing. In this work, we employ first-principles simulations to investigate the energetic stability of 16 distinct slab models representing (001), (100), (101), (110), and (111) surfaces, which are constructed by minimizing dangling bonds. Our results show that C-terminated (001) surface exhibits significantly greater stability than other surfaces under both the carbon and boron-rich conditions. Besides, we also study the defect formation energies on the C-terminated (001) surface and compare them with the cases in bulk. The high formation energies of the defects suggest a low likelihood of their occurrence on this surface, despite their formation energies being lower compared to bulk cases. Furthermore, mid-gap surface states are revealed for the top atomic layers of the C-terminated (001) surface, which are deduced at the deeper layers, and the band structures of the middle layers of this slab recover to the bulk band gap. These surface mid-gap states allow electron excitation from the valence band to these states, resulting in a reduced optical band gap compared to the bulk band gap of B<sub>4</sub>C. This provides a plausible explanation for the significantly smaller band gap observed in experiments compared to the larger gap predicted by theoretical models. Our study not only sheds light on the surface properties of B<sub>4</sub>C but also lays the groundwork for advancing this material for more advanced nuclear applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155465"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537320","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 : 2024-10-18DOI: 10.1016/j.jnucmat.2024.155463
Eun-Young Choi , Seungwoo Paek , Taehyoung Kim , In-Ho Jung , Seol Kim , Sang-Eun Bae , Jae Soo Ryu
Uranium (III) chloride (UCl3) is a crucial component of a potent nuclear recycling technology—pyroprocessing—and next-generation molten salt reactors. It is usually synthesized by reacting metallic uranium with chlorinating agents (e.g., CdCl2 and PbCl2) in molten chloride salts. In this study, we report the unexpected formation of UCl3 from metallic simulated fuel (simfuel) immersed in impure molten LiCl–KCl salt (in the presence of a small amount of residual H2O) in a stainless-steel (SS) crucible, without a chlorinating agent. We investigated various factors influencing UCl3 formation, including fuel type (metallic simfuel, pure U, oxide simfuel, or no fuel), crucible material (SS or alumina), salt composition (LiCl–KCl or LiCl), temperature (773 K or 923 K), and contact between fuel and SS crucible. UCl3 only formed when metallic fuels (simfuel or pure U) were immersed in molten salt in the SS crucible, with higher concentrations at elevated temperatures. Oxide fuels did not produce UCl3, nor did contact with the crucible affect formation. Our findings suggest that impurities, particularly moisture in the salt, corroded the SS crucible, releasing iron and chromium chlorides that reacted with metallic U to form UCl3. UCl3 formation was more pronounced in LiCl–KCl than in LiCl, and thermodynamic calculations helped establish the mechanism.
氯化铀（III）（UCl3）是一种有效的核回收技术--热处理和下一代熔盐反应堆的重要组成部分。它通常是通过金属铀与氯化剂（如 CdCl2 和 PbCl2）在熔融氯化盐中发生反应而合成的。在本研究中，我们报告了金属模拟燃料（simfuel）在不锈钢（SS）坩埚中浸入不纯的熔融 LiCl-KCl 盐（存在少量残余 H2O）后，在没有氯化剂的情况下意外地形成了 UCl3。我们研究了影响 UCl3 形成的各种因素，包括燃料类型（金属模拟燃料、纯 U、氧化物模拟燃料或无燃料）、坩埚材料（SS 或氧化铝）、盐成分（LiCl-KCl 或 LiCl）、温度（773 K 或 923 K）以及燃料与 SS 坩埚之间的接触。只有当金属燃料（simfuel 或纯 U）浸入 SS 坩埚中的熔盐时才会形成 UCl3，温度升高时浓度更高。氧化物燃料不会产生三氯化铀，与坩埚的接触也不会影响三氯化铀的形成。我们的研究结果表明，杂质，尤其是盐中的水分，腐蚀了 SS 坩埚，释放出铁和铬的氯化物，与金属铀反应生成三氯化铀。UCl3 在 LiCl-KCl 中的形成比在 LiCl 中更明显，热力学计算有助于确定其机理。
{"title":"Corrosion-mediated production of uranium(III) chloride from metallic uranium in molten LiCl–KCl salt contained within a stainless-steel crucible","authors":"Eun-Young Choi , Seungwoo Paek , Taehyoung Kim , In-Ho Jung , Seol Kim , Sang-Eun Bae , Jae Soo Ryu","doi":"10.1016/j.jnucmat.2024.155463","DOIUrl":"10.1016/j.jnucmat.2024.155463","url":null,"abstract":"<div><div>Uranium (III) chloride (UCl<sub>3</sub>) is a crucial component of a potent nuclear recycling technology—pyroprocessing—and next-generation molten salt reactors. It is usually synthesized by reacting metallic uranium with chlorinating agents (e.g., CdCl<sub>2</sub> and PbCl<sub>2</sub>) in molten chloride salts. In this study, we report the unexpected formation of UCl<sub>3</sub> from metallic simulated fuel (simfuel) immersed in impure molten LiCl–KCl salt (in the presence of a small amount of residual H<sub>2</sub>O) in a stainless-steel (SS) crucible, without a chlorinating agent. We investigated various factors influencing UCl<sub>3</sub> formation, including fuel type (metallic simfuel, pure U, oxide simfuel, or no fuel), crucible material (SS or alumina), salt composition (LiCl–KCl or LiCl), temperature (773 K or 923 K), and contact between fuel and SS crucible. UCl<sub>3</sub> only formed when metallic fuels (simfuel or pure U) were immersed in molten salt in the SS crucible, with higher concentrations at elevated temperatures. Oxide fuels did not produce UCl<sub>3</sub>, nor did contact with the crucible affect formation. Our findings suggest that impurities, particularly moisture in the salt, corroded the SS crucible, releasing iron and chromium chlorides that reacted with metallic U to form UCl<sub>3</sub>. UCl<sub>3</sub> formation was more pronounced in LiCl–KCl than in LiCl, and thermodynamic calculations helped establish the mechanism.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155463"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537328","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}
Fretting corrosion of stainless steel in the LBE affects the safety of lead-cooled fast reactors. Slip amplitude and normal load are the main mechanical factors affecting fretting wear behavior. Thus, the damage mechanism of 316L stainless steel at 350 °C LBE influenced by slip amplitude and normal load was investigated by jointly utilizing multiple characterization methods. The results indicate that the normal load and slip amplitude essentially affect the tangential stress and relative sliding value in the contact area, leading to different slip regions and damage mechanisms. In the mixed slip region, the damage mechanism is adhesion and delamination cracks. The increase in tangential stress leads to decrease in relative sliding. The thick wear debris layer attached to the worn surface can protect the substrate from being attacked by the LBE. In the gross slip region, the damage mechanism is abrasive wear and dissolution corrosion. The increase in relative sliding causes more damage and Ni dissolution, leading to the transformation from austenite to ferrite and internal strain, making the substrate more susceptible to damage and increasing the risk of liquid metal embrittlement (LME) of austenitic stainless steel at 350 °C. Accordingly, a model for different damage mechanisms was proposed. These results can provide important information on the fretting damage related to the LBE environment.
铅冷快堆中不锈钢的烧蚀会影响铅冷快堆的安全。滑动幅度和法向载荷是影响烧蚀磨损行为的主要机械因素。因此，通过联合使用多种表征方法，研究了 316L 不锈钢在 350 °C 铅冷快堆中受滑动幅度和法向载荷影响的损伤机理。结果表明，法向载荷和滑移振幅主要影响接触区的切向应力和相对滑移值，从而导致不同的滑移区域和损伤机制。在混合滑移区域，损坏机制是粘着和分层裂纹。切向应力的增加导致相对滑动的减小。磨损表面附着的厚磨损碎屑层可以保护基体免受 LBE 的侵蚀。在粗滑动区域，破坏机制是磨料磨损和溶解腐蚀。相对滑动的增加会造成更多的损伤和镍的溶解，导致奥氏体向铁素体的转变和内部应变，使基体更容易受到损伤，增加了奥氏体不锈钢在 350 °C 下发生液态金属脆性（LME）的风险。因此，我们提出了不同损伤机制的模型。这些结果可为与 LBE 环境相关的烧蚀损伤提供重要信息。
{"title":"A study on the fretting corrosion of 316L in static lead-bismuth eutectic (LBE): The role of slip amplitude and normal force on damage mechanism at 350 °C","authors":"Hui Chen , Wenjie Pei , Shengzan Zhang , Wei Tan , Guorui Zhu","doi":"10.1016/j.jnucmat.2024.155466","DOIUrl":"10.1016/j.jnucmat.2024.155466","url":null,"abstract":"<div><div>Fretting corrosion of stainless steel in the LBE affects the safety of lead-cooled fast reactors. Slip amplitude and normal load are the main mechanical factors affecting fretting wear behavior. Thus, the damage mechanism of 316L stainless steel at 350 °C LBE influenced by slip amplitude and normal load was investigated by jointly utilizing multiple characterization methods. The results indicate that the normal load and slip amplitude essentially affect the tangential stress and relative sliding value in the contact area, leading to different slip regions and damage mechanisms. In the mixed slip region, the damage mechanism is adhesion and delamination cracks. The increase in tangential stress leads to decrease in relative sliding. The thick wear debris layer attached to the worn surface can protect the substrate from being attacked by the LBE. In the gross slip region, the damage mechanism is abrasive wear and dissolution corrosion. The increase in relative sliding causes more damage and Ni dissolution, leading to the transformation from austenite to ferrite and internal strain, making the substrate more susceptible to damage and increasing the risk of liquid metal embrittlement (LME) of austenitic stainless steel at 350 °C. Accordingly, a model for different damage mechanisms was proposed. These results can provide important information on the fretting damage related to the LBE environment.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155466"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142553738","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 : 2024-10-17DOI: 10.1016/j.jnucmat.2024.155462
A.M. Kpemou , S. Guilbert , J. Desquines , T. Taurines , M.C. Baietto , B. Normand , J. Soulacroix , A. Ambard , F. Bourlier
The focus of this study is about a new experimental approach for a separate effects study of the secondary hydriding phenomenon under LOCA conditions. Many nuclear institutes perform semi–integrals tests to study the cladding behaviour during a LOCA transient. Those tests combined several phenomena and performing a detailed analysis of the secondary hydriding phenomenon using these tests can be challenging. A dedicated experimental protocol aiming at simulating secondary hydriding has been set up. Separate effects tests (SETs) were then carried out using this protocol to study the effects of both oxidation duration and temperature, on the hydrogen absorption during the oxidation stage of the LOCA transient on M5Framatome1 cladding. The effects of gap size were also investigated. Metallographic analysis has been used to characterise the M5Framatome clad metallurgical transformation after the high–temperature (HT) oxidation.
{"title":"A new approach to investigate secondary hydriding phenomenon on M5Framatome clads under high–temperature LOCA conditions","authors":"A.M. Kpemou , S. Guilbert , J. Desquines , T. Taurines , M.C. Baietto , B. Normand , J. Soulacroix , A. Ambard , F. Bourlier","doi":"10.1016/j.jnucmat.2024.155462","DOIUrl":"10.1016/j.jnucmat.2024.155462","url":null,"abstract":"<div><div>The focus of this study is about a new experimental approach for a separate effects study of the secondary hydriding phenomenon under LOCA conditions. Many nuclear institutes perform semi–integrals tests to study the cladding behaviour during a LOCA transient. Those tests combined several phenomena and performing a detailed analysis of the secondary hydriding phenomenon using these tests can be challenging. A dedicated experimental protocol aiming at simulating secondary hydriding has been set up. Separate effects tests (SETs) were then carried out using this protocol to study the effects of both oxidation duration and temperature, on the hydrogen absorption during the oxidation stage of the LOCA transient on M5<sub>Framatome</sub><span><span><sup>1</sup></span></span> cladding. The effects of gap size were also investigated. Metallographic analysis has been used to characterise the M5<sub>Framatome</sub> clad metallurgical transformation after the high–temperature (HT) oxidation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155462"},"PeriodicalIF":2.8,"publicationDate":"2024-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537326","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 : 2024-10-16DOI: 10.1016/j.jnucmat.2024.155460
X.W. Zhou, M.E. Foster
Understanding and containing hydrogen isotope diffusion is crucial for many nuclear applications. In situ experiments have consistently shown that radiation significantly enhances isotope diffusion in austenitic stainless steels. Despite extensive research, the mechanism behind this phenomenon remains elusive, as most radiation-induced defects (e.g., vacancies, dislocations, and grain boundaries) typically trap hydrogen, thereby slowing diffusion. While grain boundaries may increase in-plane diffusivity and interstitials may enhance diffusion due to material swelling, these effects are relatively minor. Utilizing an Fe-Ni-Cr-H interatomic potential for stainless steels, we conducted extensive molecular dynamics simulations to investigate the origins of radiation-enhanced diffusion. Our findings reveal that when a system is resolidified, mimicking defects created by radiation displacements, the resulting structure contains a mixture of phases, boundaries, and dislocation networks. This defective structure significantly increases hydrogen diffusivity, enhancing it by approximately 1.7 times at 900 K. These results suggest that the complex defect structures formed during radiation displacements are the primary drivers of the observed diffusion enhancement, providing valuable insights into the mechanisms underlying radiation-enhanced diffusion in nuclear materials.
{"title":"Demystify radiation-enhanced hydrogen isotope diffusion in Fe-Ni-Cr austenitic stainless steels","authors":"X.W. Zhou, M.E. Foster","doi":"10.1016/j.jnucmat.2024.155460","DOIUrl":"10.1016/j.jnucmat.2024.155460","url":null,"abstract":"<div><div>Understanding and containing hydrogen isotope diffusion is crucial for many nuclear applications. In situ experiments have consistently shown that radiation significantly enhances isotope diffusion in austenitic stainless steels. Despite extensive research, the mechanism behind this phenomenon remains elusive, as most radiation-induced defects (e.g., vacancies, dislocations, and grain boundaries) typically trap hydrogen, thereby slowing diffusion. While grain boundaries may increase in-plane diffusivity and interstitials may enhance diffusion due to material swelling, these effects are relatively minor. Utilizing an Fe-Ni-Cr-H interatomic potential for stainless steels, we conducted extensive molecular dynamics simulations to investigate the origins of radiation-enhanced diffusion. Our findings reveal that when a system is resolidified, mimicking defects created by radiation displacements, the resulting structure contains a mixture of phases, boundaries, and dislocation networks. This defective structure significantly increases hydrogen diffusivity, enhancing it by approximately 1.7 times at 900 K. These results suggest that the complex defect structures formed during radiation displacements are the primary drivers of the observed diffusion enhancement, providing valuable insights into the mechanisms underlying radiation-enhanced diffusion in nuclear materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155460"},"PeriodicalIF":2.8,"publicationDate":"2024-10-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537319","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}
FeCrAl coatings were applied to the surface of F/M steel using ultrasonic vibration-assisted laser cladding (UVALC) technique. The introduction of an ultrasonic field refined the microstructure of the FeCrAl coating, enhancing its microhardness and the integrity of the oxide film at elevated temperatures. The increased hardness led to a shift in the wear mechanism from oxidation wear to abrasive wear. In high-temperature conditions, a finer microstructure of the coating resulted in a denser oxide layer, improving the tribological properties and oxidation resistance of the coating. Furthermore, high-temperature oxidation analysis revealed that the predominant oxides formed were Fe2O3 and Cr2O3.
{"title":"Effect of ultrasonic field on the friction and oxidation characteristics of FeCrAl coatings","authors":"Changhao Liu, Xiufang Cui, Guo Jin, Meng Qi, Jiaxin Zhao, Di Wu, Xin Wen","doi":"10.1016/j.jnucmat.2024.155457","DOIUrl":"10.1016/j.jnucmat.2024.155457","url":null,"abstract":"<div><div>FeCrAl coatings were applied to the surface of F/M steel using ultrasonic vibration-assisted laser cladding (UVALC) technique. The introduction of an ultrasonic field refined the microstructure of the FeCrAl coating, enhancing its microhardness and the integrity of the oxide film at elevated temperatures. The increased hardness led to a shift in the wear mechanism from oxidation wear to abrasive wear. In high-temperature conditions, a finer microstructure of the coating resulted in a denser oxide layer, improving the tribological properties and oxidation resistance of the coating. Furthermore, high-temperature oxidation analysis revealed that the predominant oxides formed were Fe<sub>2</sub>O<sub>3</sub> and Cr<sub>2</sub>O<sub>3</sub>.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155457"},"PeriodicalIF":2.8,"publicationDate":"2024-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537318","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 : 2024-10-13DOI: 10.1016/j.jnucmat.2024.155458
Wen Wang , Liujie Yang , Hongchen Qian , Zhaoguang Zhu , Guangjian Zhu , Jibo Tan , Jinyang Huang , Jintao Lu , Wenjun Kuang
A FeAl coating was fabricated on T91 steel via slurry aluminizing. The corrosion behavior of coated and uncoated samples was assessed in static LBE (lead-bismuth eutectic) with two both high and low dissolved oxygen concentrations at 550 °C. The coating was mostly intact and exhibited great corrosion resistance compared to the uncoated specimens regardless of the oxygen concentration. That is because the coating can form a protective alumina film on the surface at extremely low dissolved oxygen level. This coating is of significant engineering value in enhancing the corrosion resistance of Fe base alloy in LBE.
通过浆状镀铝在 T91 钢上制作了铁铝涂层。在 550 °C 的静态 LBE（铅铋共晶）条件下，评估了有涂层和无涂层试样的腐蚀行为。与无涂层试样相比，无论氧气浓度如何，涂层大部分都完好无损，并表现出很强的耐腐蚀性。这是因为涂层能在极低的溶解氧水平下在表面形成一层氧化铝保护膜。这种涂层在提高铁基合金在 LBE 中的耐腐蚀性方面具有重要的工程价值。
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To understand the core degradation process at the Fukushima Daiichi Nuclear Power Station, the oxidation of boron carbide–stainless steel alloy under steam starvation condition was studied at temperatures in the range of 1,288–1,573 K. Low steam supply led to swift Fe–O layer formation, embedding Fe–B–O and Fe–Cr–O, and boron evaporation mainly as oxides was observed through the Fe–B–O phase precipitated in the Fe–O layer. The rate constant of boron evaporation kB was derived from the measured data as kB = 0.0157 exp (–79.8 × 103/RT) for T ≥ 1,423 K and kB = 8.69 × 10−5exp (–44.4 × 103/RT) for T < 1,423 K where R and T are the gas constant and temperature, respectively. The obtained constant was comparable to the reaction rate of B4C oxidation. In addition, a test with an even more decreased steam supply was conducted to examine the impact of steam quantity on the boron evaporation kinetics. Consequently, it was confirmed that decreasing the oxygen supply resulted in a slowdown of outer Fe–O layer formation, which enhances the outwards diffusion of B and allows greater evaporation of B oxides.
为了解福岛第一核电站堆芯降解过程，我们在 1288-1,573 K 的温度范围内研究了碳化硼-不锈钢合金在蒸汽饥饿条件下的氧化过程。低蒸汽供应导致 Fe-O 层迅速形成，嵌入了 Fe-B-O 和 Fe-Cr-O，并通过 Fe-O 层中析出的 Fe-B-O 相观察到硼主要以氧化物的形式蒸发。根据测量数据得出硼蒸发的速率常数 kB：T ≥ 1,423 K 时，kB = 0.0157 exp (-79.8 × 103/RT)；T < 1,423 K 时，kB = 8.69 × 10-5 exp (-44.4 × 103/RT)，其中 R 和 T 分别为气体常数和温度。所得常数与 B4C 氧化反应速率相当。此外，还进行了一次蒸汽供应量更小的试验，以检验蒸汽量对硼蒸发动力学的影响。结果证实，减少供氧量会导致外层 Fe-O 层的形成速度减慢，从而加强硼的向外扩散，使硼氧化物的蒸发量增加。
{"title":"Evaluation of boron evaporation kinetics from stainless-steel–B4C alloy during steam oxidation at high temperatures","authors":"Kosuke Inoue , Ayumi Itoh , Masato Mizokami , Mutsumi Hirai","doi":"10.1016/j.jnucmat.2024.155456","DOIUrl":"10.1016/j.jnucmat.2024.155456","url":null,"abstract":"<div><div>To understand the core degradation process at the Fukushima Daiichi Nuclear Power Station, the oxidation of boron carbide–stainless steel alloy under steam starvation condition was studied at temperatures in the range of 1,288–1,573 K. Low steam supply led to swift Fe–O layer formation, embedding Fe–B–O and Fe–Cr–O, and boron evaporation mainly as oxides was observed through the Fe–B–O phase precipitated in the Fe–O layer. The rate constant of boron evaporation <em>k</em><sub>B</sub> was derived from the measured data as <em>k</em><sub>B</sub> = 0.0157 <em>exp</em> (–79.8 × 10<sup>3</sup>/<em>RT</em>) for <em>T</em> ≥ 1,423 K and <em>k</em><sub>B</sub> = 8.69 × 10<sup>−5</sup> <em>exp</em> (–44.4 × 10<sup>3</sup>/<em>RT</em>) for <em>T</em> < 1,423 K where <em>R</em> and <em>T</em> are the gas constant and temperature, respectively. The obtained constant was comparable to the reaction rate of B<sub>4</sub>C oxidation. In addition, a test with an even more decreased steam supply was conducted to examine the impact of steam quantity on the boron evaporation kinetics. Consequently, it was confirmed that decreasing the oxygen supply resulted in a slowdown of outer Fe–O layer formation, which enhances the outwards diffusion of B and allows greater evaporation of B oxides.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155456"},"PeriodicalIF":2.8,"publicationDate":"2024-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537323","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}
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