This work focuses on neutron irradiated Zr-1Nb alloy, using High Resolution Transmission Electron Microscopy (HRTEM) to investigate the oxidation behavior of β-Nb at different distances from the Oxide /Metal (O/M) interface within the oxide film. Results show that β-Nb was initially oxidized to T-NbO2 at 0 nm at O/M interface, then into a complex morphology of T-NbO2, M-Nb2O5, and O-Nb2O5 within 600 nm. Finally, it was completely oxidized to M-Nb2O5 within 800 nm. β-Nb in this study did not exhibit amorphous morphology within observed distances. In addition, Inverse Fast Fourier Transformation (IFFT) and Weak Beam Dark Field (WBDF) techniques are employed to characterize the dislocation density and distribution in the oxide film, results indicate that the distribution of dislocations generated by neutron irradiation in the oxide film is relatively uniform and neutron irradiation is not the primary reason affecting the oxidation behavior of β-Nb.
{"title":"Oxidation behavior of β-Nb formed in Zr-1Nb under neutron irradiation in PWR conditions","authors":"Xue Han, Huacai Wang, Huanlin Cheng, Jinze Sun, Lina Guo, Wulin Song, Huize Fan","doi":"10.1016/j.jnucmat.2024.155478","DOIUrl":"10.1016/j.jnucmat.2024.155478","url":null,"abstract":"<div><div>This work focuses on neutron irradiated Zr-1Nb alloy, using High Resolution Transmission Electron Microscopy (HRTEM) to investigate the oxidation behavior of β-Nb at different distances from the Oxide /Metal (O/M) interface within the oxide film. Results show that β-Nb was initially oxidized to T-NbO<sub>2</sub> at 0 nm at O/M interface, then into a complex morphology of T-NbO<sub>2</sub>, M-Nb<sub>2</sub>O<sub>5</sub>, and O-Nb<sub>2</sub>O<sub>5</sub> within 600 nm. Finally, it was completely oxidized to M-Nb<sub>2</sub>O<sub>5</sub> within 800 nm. β-Nb in this study did not exhibit amorphous morphology within observed distances. In addition, Inverse Fast Fourier Transformation (IFFT) and Weak Beam Dark Field (WBDF) techniques are employed to characterize the dislocation density and distribution in the oxide film, results indicate that the distribution of dislocations generated by neutron irradiation in the oxide film is relatively uniform and neutron irradiation is not the primary reason affecting the oxidation behavior of β-Nb.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155478"},"PeriodicalIF":2.8,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652826","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}
The deterministic approach widely adopted in the design of structural components relies on systematically defined design limits using empirically determined safety factors. However, this approach is not always appropriate because structures are subjected to a variety of loads in the practical environment, which may result in excessively conservative design limits. In recent years, a more rigorous probabilistic approach that incorporates material strength distributions has become an important solution. In the probabilistic approach, the probability density functions of material strength properties underpin the design criteria. The objective of this study is to identify the density distribution functions that best describe tensile properties of irradiated F82H to define a reference strength for DEMO design. Due to the limited number of existing data, this study specifically employs a Bayesian prediction method based on Monte Carlo simulations to determine a material reference value with statistical reliability and to investigate its effectiveness. For example, the dependence of tensile properties of 300 °C irradiated materials on irradiation damage and the range predicted by 95% Bayesian estimation was evaluated. As a statistical model for the dose dependence of statistical parameters, the normal distribution exhibited a better fit for 0.2% proof strength and tensile strength, whereas the distribution of total elongation data gave comparable reference values for both the normal and Weibull distribution models. Both models gave comparable criteria for the distribution of total elongation data. The Weibull model also gave better results for uniform elongation. The function best describing the model was a logarithmic law for both 0.2% proof strength and tensile strength, while a power law for both total and uniform elongation, which allowed for more comprehensive data prediction of irradiation data with statistical accuracy for DEMO reactor design.
{"title":"Determining reference standard strength for neutron-irradiated reduced activation ferritic/martensitic steel F82H by Bayesian method","authors":"Takashi Nozawa , Hiroyasu Tanigawa , Taichiro Kato , Xiang (Frank) Chen , Yutai Katoh","doi":"10.1016/j.jnucmat.2024.155486","DOIUrl":"10.1016/j.jnucmat.2024.155486","url":null,"abstract":"<div><div>The deterministic approach widely adopted in the design of structural components relies on systematically defined design limits using empirically determined safety factors. However, this approach is not always appropriate because structures are subjected to a variety of loads in the practical environment, which may result in excessively conservative design limits. In recent years, a more rigorous probabilistic approach that incorporates material strength distributions has become an important solution. In the probabilistic approach, the probability density functions of material strength properties underpin the design criteria. The objective of this study is to identify the density distribution functions that best describe tensile properties of irradiated F82H to define a reference strength for DEMO design. Due to the limited number of existing data, this study specifically employs a Bayesian prediction method based on Monte Carlo simulations to determine a material reference value with statistical reliability and to investigate its effectiveness. For example, the dependence of tensile properties of 300 °C irradiated materials on irradiation damage and the range predicted by 95% Bayesian estimation was evaluated. As a statistical model for the dose dependence of statistical parameters, the normal distribution exhibited a better fit for 0.2% proof strength and tensile strength, whereas the distribution of total elongation data gave comparable reference values for both the normal and Weibull distribution models. Both models gave comparable criteria for the distribution of total elongation data. The Weibull model also gave better results for uniform elongation. The function best describing the model was a logarithmic law for both 0.2% proof strength and tensile strength, while a power law for both total and uniform elongation, which allowed for more comprehensive data prediction of irradiation data with statistical accuracy for DEMO reactor design.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155486"},"PeriodicalIF":2.8,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571578","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 : 2024-10-29DOI: 10.1016/j.jnucmat.2024.155487
S. Mondal , M. Sen , S.K. Makineni , P. Ghosh , A. Sarkar , R. Kapoor , S. Suwas
In this investigation, the effect of 5.6 MeV proton irradiation on the microstructure and mechanical properties of coarse grained (CG) and nanocrystalline (NC) Nb-1wt.%Zr (NZ) has been analysed. Bulk nanocrystalline microstructure was obtained by subjecting the alloy to room temperature high pressure torsion under 6 GPa hydrostatic pressure and 5 rotations. The CG and NC samples were irradiated at doses of 1.9 × 1017 p/cm2 and 1.8 × 1017 p/cm2, respectively. Microstructural parameters like crystallite size, dislocation density, and dislocation arrangements were studied in detail using X-ray line profile analysis (XLPA) by Convolutional Multiple Whole Profile (CMWP) fitting. Microscopic observations were made with electron microscopy techniques in the scanning and transmission modes. Differential Scanning Calorimetry (DSC) was performed to estimate the concentration of vacancies after HPT processing and irradiation. Tensile tests of irradiated CG and NC irradiated samples were performed and compared to those in unirradiated conditions. In the NC condition, not only did the irradiated sample show higher ultimate tensile strength but also twice the amount of uniform elongation as compared to the irradiated CG sample. The fracture surface clearly exhibited this higher plasticity post-irradiation in the NC samples. The change in deformation mechanisms due to nano-structuring of the microstructure has been anticipated to be a reason for the increase in ductility in a single-phase alloy has been explained thereafter.
{"title":"Role of severe plastic deformation on mechanical behavior of irradiated materials: A case study with Nb-1Zr alloy","authors":"S. Mondal , M. Sen , S.K. Makineni , P. Ghosh , A. Sarkar , R. Kapoor , S. Suwas","doi":"10.1016/j.jnucmat.2024.155487","DOIUrl":"10.1016/j.jnucmat.2024.155487","url":null,"abstract":"<div><div>In this investigation, the effect of 5.6 MeV proton irradiation on the microstructure and mechanical properties of coarse grained (CG) and nanocrystalline (NC) Nb-1wt.%Zr (NZ) has been analysed. Bulk nanocrystalline microstructure was obtained by subjecting the alloy to room temperature high pressure torsion under 6 GPa hydrostatic pressure and 5 rotations. The CG and NC samples were irradiated at doses of 1.9 × 10<sup>17</sup> p/cm<sup>2</sup> and 1.8 × 10<sup>17</sup> p/cm<sup>2</sup>, respectively. Microstructural parameters like crystallite size, dislocation density, and dislocation arrangements were studied in detail using X-ray line profile analysis (XLPA) by Convolutional Multiple Whole Profile (CMWP) fitting. Microscopic observations were made with electron microscopy techniques in the scanning and transmission modes. Differential Scanning Calorimetry (DSC) was performed to estimate the concentration of vacancies after HPT processing and irradiation. Tensile tests of irradiated CG and NC irradiated samples were performed and compared to those in unirradiated conditions. In the NC condition, not only did the irradiated sample show higher ultimate tensile strength but also twice the amount of uniform elongation as compared to the irradiated CG sample. The fracture surface clearly exhibited this higher plasticity post-irradiation in the NC samples. The change in deformation mechanisms due to nano-structuring of the microstructure has been anticipated to be a reason for the increase in ductility in a single-phase alloy has been explained thereafter.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155487"},"PeriodicalIF":2.8,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652815","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 : 2024-10-28DOI: 10.1016/j.jnucmat.2024.155485
Sitendu Mandal , Gattu Suneel , Jayaprakasam Selvakumar , Kaushik Biswas , Srikrishna Manna , Sourav Nag , Balram Ambade
High-level radioactive liquid waste (HLW) is immobilized in a glass matrix through a process called vitrification. In this process, HLW and glass-forming oxides are combined in a pre-determined ratio within a glass melter to produce a vitrified waste form. The properties of this waste form, including its ability to accommodate different radioactive isotopes, depend on the composition of the base glass.
In the present study, multi-component amorphous borosilicate-based glasses (SiO2-B2O3-Na2O-TiO2-Fe2O3-CaO-K2O) in bead form (diameter 2–3 mm) were developed. The elemental composition of the glass beads (GBs) was analyzed using an optical emission spectrometer. Additionally, the GBs underwent various physico-chemical analyses, including functional group identification, thermal, electrical, and mechanical properties, as well as viscosity and chemical durability assessments, to identify the optimal glass compositions. The influence of Na2O on the pouring temperature was also examined. Crushing strength and attrition rate measurements were conducted to confirm the suitability of GBs for remote feeding into the melter. The GBs developed in the study are unique, with significant potential for worldwide use in vitrification facilities, particularly in continuous vitrification systems employing Joule Heated Ceramic Melter (JHCM) technology.
{"title":"Synthesis and characterization of multi-component borosilicate glass beads for radioactive liquid waste immobilisation","authors":"Sitendu Mandal , Gattu Suneel , Jayaprakasam Selvakumar , Kaushik Biswas , Srikrishna Manna , Sourav Nag , Balram Ambade","doi":"10.1016/j.jnucmat.2024.155485","DOIUrl":"10.1016/j.jnucmat.2024.155485","url":null,"abstract":"<div><div>High-level radioactive liquid waste (HLW) is immobilized in a glass matrix through a process called vitrification. In this process, HLW and glass-forming oxides are combined in a pre-determined ratio within a glass melter to produce a vitrified waste form. The properties of this waste form, including its ability to accommodate different radioactive isotopes, depend on the composition of the base glass.</div><div>In the present study, multi-component amorphous borosilicate-based glasses (SiO<sub>2</sub>-B<sub>2</sub>O<sub>3</sub>-Na<sub>2</sub>O-TiO<sub>2</sub>-Fe<sub>2</sub>O<sub>3</sub>-CaO-K<sub>2</sub>O) in bead form (diameter 2–3 mm) were developed. The elemental composition of the glass beads (GBs) was analyzed using an optical emission spectrometer. Additionally, the GBs underwent various physico-chemical analyses, including functional group identification, thermal, electrical, and mechanical properties, as well as viscosity and chemical durability assessments, to identify the optimal glass compositions. The influence of Na<sub>2</sub>O on the pouring temperature was also examined. Crushing strength and attrition rate measurements were conducted to confirm the suitability of GBs for remote feeding into the melter. The GBs developed in the study are unique, with significant potential for worldwide use in vitrification facilities, particularly in continuous vitrification systems employing Joule Heated Ceramic Melter (JHCM) technology.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155485"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652878","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 : 2024-10-28DOI: 10.1016/j.jnucmat.2024.155469
S. Lam , D. Frazer , F. Cappia , M. Nelson , S. Samuha , S. Pitts , B. Harris , P. Hosemann
Zircaloy-4 is an essential material for cladding structures within fission-based reactors. To explore the changes in properties measured on differing length scales, FIB-machined micro-scale tensile tests were performed on both irradiated and control groups of Zircaloy-4. This was correlated with tensile testing on femtosecond laser-machined meso‑scale specimens. Pronounced size effects were found when varying specimen geometry. Increases in tensile geometry size were associated with a reduction in measured yield stress for both irradiated and unirradiated samples. Meso-scale testing found strength and strain values similar to that of bulk-scale testing.
{"title":"Length scale effects of micro- and meso‑scale tensile tests of unirradiated and irradiated Zircaloy-4 cladding","authors":"S. Lam , D. Frazer , F. Cappia , M. Nelson , S. Samuha , S. Pitts , B. Harris , P. Hosemann","doi":"10.1016/j.jnucmat.2024.155469","DOIUrl":"10.1016/j.jnucmat.2024.155469","url":null,"abstract":"<div><div>Zircaloy-4 is an essential material for cladding structures within fission-based reactors. To explore the changes in properties measured on differing length scales, FIB-machined micro-scale tensile tests were performed on both irradiated and control groups of Zircaloy-4. This was correlated with tensile testing on femtosecond laser-machined meso‑scale specimens. Pronounced size effects were found when varying specimen geometry. Increases in tensile geometry size were associated with a reduction in measured yield stress for both irradiated and unirradiated samples. Meso-scale testing found strength and strain values similar to that of bulk-scale testing.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155469"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571579","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 : 2024-10-28DOI: 10.1016/j.jnucmat.2024.155484
Tian-Xing Yang, Peng Dou, Chang-Jun Zhou
FeCrAl oxide dispersion strengthened (ODS) steel is one of the most promising candidate cladding materials in generation IV nuclear reactors due to its exceptional macro-properties. To address the stringent performance requirements of supercritical water-cooled reactors (SCPWRs), two FeCrAl-ODS steels, i.e., 3Al–0.1Ti (Fe–16Cr–3Al–0.1Ti–0.34Y2O3) and 2Al–0.1Ti–0.35Ce (Fe–16Cr–2Al–0.1Ti–0.35Ce–0.36Y2O3), were developed. This study aims to investigate how Ce addition influences the microstructure and the formation mechanisms of various oxides in ODS steels. Therefore, the grain & nanoparticle morphologies, and crystal & interface structures of nano-scale oxides of the two ODS steels were studied by transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM) and high-resolution transmission electron microscopy (HRTEM). The mean grain diameter of 3Al–0.1Ti and 2Al–0.1Ti–0.35Ce is 1.1 μm and 0.82 μm, respectively. Compared with 3Al–0.1Ti, the average diameter of particles of 2Al–0.1Ti–0.35Ce is relatively smaller. The results indicate that adding Ce can refine the grains and nano-sized particles. For 3Al–0.1Ti, the main particles are Y–Al–O with a proportion of ∼81.4 %. For 2Al–0.1Ti–0.35Ce, the main particles are Y–Ce and Y–Ti oxides with quantity ratios of ∼52.2 % and ∼22.1 %, respectively, while the quantity ratio of Y–Al oxides is only 12.3 %. This indicates that adding Ce can impede the occurrence of Y–Al–O while facilitating the generation of Y–Ce–O. Moreover, it is the first time that Y2Ce2O7 oxide has been detected in yttria-added ODS steels with Ce. The findings obtained from this study provide key insights into the mechanisms of oxide formation & polymorphic transitions, and microstructural differences due to Ce addition. This will provide pivotal direction for the optimization of alloy compositions, promoting the innovation of ODS steels. Additionally, the feasibility analysis of the two ODS steels indicates their applicability to the SCPWR fuel cladding.
{"title":"Effects of Ce addition on the morphology, crystal and metal/oxide interface structures of nanoparticles in FeCrAl-ODS steels","authors":"Tian-Xing Yang, Peng Dou, Chang-Jun Zhou","doi":"10.1016/j.jnucmat.2024.155484","DOIUrl":"10.1016/j.jnucmat.2024.155484","url":null,"abstract":"<div><div>FeCrAl oxide dispersion strengthened (ODS) steel is one of the most promising candidate cladding materials in generation IV nuclear reactors due to its exceptional macro-properties. To address the stringent performance requirements of supercritical water-cooled reactors (SCPWRs), two FeCrAl-ODS steels, i.e., 3Al–0.1Ti (Fe–16Cr–3Al–0.1Ti–0.34Y<sub>2</sub>O<sub>3</sub>) and 2Al–0.1Ti–0.35Ce (Fe–16Cr–2Al–0.1Ti–0.35Ce–0.36Y<sub>2</sub>O<sub>3</sub>), were developed. This study aims to investigate how Ce addition influences the microstructure and the formation mechanisms of various oxides in ODS steels. Therefore, the grain & nanoparticle morphologies, and crystal & interface structures of nano-scale oxides of the two ODS steels were studied by transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM) and high-resolution transmission electron microscopy (HRTEM). The mean grain diameter of 3Al–0.1Ti and 2Al–0.1Ti–0.35Ce is 1.1 μm and 0.82 μm, respectively. Compared with 3Al–0.1Ti, the average diameter of particles of 2Al–0.1Ti–0.35Ce is relatively smaller. The results indicate that adding Ce can refine the grains and nano-sized particles. For 3Al–0.1Ti, the main particles are Y–Al–O with a proportion of ∼81.4 %. For 2Al–0.1Ti–0.35Ce, the main particles are Y–Ce and Y–Ti oxides with quantity ratios of ∼52.2 % and ∼22.1 %, respectively, while the quantity ratio of Y–Al oxides is only 12.3 %. This indicates that adding Ce can impede the occurrence of Y–Al–O while facilitating the generation of Y–Ce–O. Moreover, it is the first time that Y<sub>2</sub>Ce<sub>2</sub>O<sub>7</sub> oxide has been detected in yttria-added ODS steels with Ce. The findings obtained from this study provide key insights into the mechanisms of oxide formation & polymorphic transitions, and microstructural differences due to Ce addition. This will provide pivotal direction for the optimization of alloy compositions, promoting the innovation of ODS steels. Additionally, the feasibility analysis of the two ODS steels indicates their applicability to the SCPWR fuel cladding.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155484"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652828","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 : 2024-10-28DOI: 10.1016/j.jnucmat.2024.155476
Sarah A. Khan , Jason L. Schulthess , Indrajit Charit , Aaron Craft , William Chuirazzi , Jatuporn Burns , David Frazer , Nicolas Woolstenhulme , Robert O'Brien
The National Aeronautics and Space Administration's return to space nuclear propulsion stems from the need for a more efficient method of space travel. Nuclear thermal propulsion systems have been shown to be two times more efficient than chemical propulsion. NASA's Sirius program was created to fabricate and test fuels for space nuclear propulsion, specifically to determine their performance under prototypical startup conditions. The Sirius project featured 4 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A, -2B, and -3 featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). This study discusses the Sirius-2A and -2B irradiation experiments at the Idaho National Laboratory, specifically their performance under irradiation at the Transient Reactor Test Facility. It was found that the fuel samples overall did not exhibit significant cracking, though the Sirius-2A fuel did have one large crack on the surface of the fuel. There was minimal hydrogen absorption in the samples, though it is unknown if the absorption occurred during irradiation or during fabrication. Mechanical testing indicated that the UN fuel demonstrated ceramic behavior as expected, and the Mo/W matrix demonstrated linear elastic behavior to failure.
{"title":"Post-irradiation examination of UN-Mo-W fuels for space nuclear propulsion","authors":"Sarah A. Khan , Jason L. Schulthess , Indrajit Charit , Aaron Craft , William Chuirazzi , Jatuporn Burns , David Frazer , Nicolas Woolstenhulme , Robert O'Brien","doi":"10.1016/j.jnucmat.2024.155476","DOIUrl":"10.1016/j.jnucmat.2024.155476","url":null,"abstract":"<div><div>The National Aeronautics and Space Administration's return to space nuclear propulsion stems from the need for a more efficient method of space travel. Nuclear thermal propulsion systems have been shown to be two times more efficient than chemical propulsion. NASA's Sirius program was created to fabricate and test fuels for space nuclear propulsion, specifically to determine their performance under prototypical startup conditions. The Sirius project featured 4 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A, -2B, and -3 featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). This study discusses the Sirius-2A and -2B irradiation experiments at the Idaho National Laboratory, specifically their performance under irradiation at the Transient Reactor Test Facility. It was found that the fuel samples overall did not exhibit significant cracking, though the Sirius-2A fuel did have one large crack on the surface of the fuel. There was minimal hydrogen absorption in the samples, though it is unknown if the absorption occurred during irradiation or during fabrication. Mechanical testing indicated that the UN fuel demonstrated ceramic behavior as expected, and the Mo/W matrix demonstrated linear elastic behavior to failure.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155476"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552505","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 : 2024-10-28DOI: 10.1016/j.jnucmat.2024.155483
Ilia V. Voronov , Vladislav S. Nikolaev , Alexey V. Timofeev , Vladimir V. Stegailov
A bcc iron bicrystal in contact with liquid lead is studied in molecular dynamics simulations to describe the atomistic mechanism of liquid metal corrosion in the activation controlled case. In this process, the main structural features involved are Fe grain boundaries and Fe-Pb interfaces. The atomistic model considered reveals that the interplay of atomic processes such as surface self-diffusion of Fe and dissolution of Fe into Pb determines the mechanism and kinetics law of liquid metal corrosion. Analysis of the proposed mechanism explains the dependence between the kinetics of liquid metal corrosion and the grain size of the specimen.
{"title":"Atomistic mechanism of activation controlled liquid metal corrosion at the Fe-Pb interface","authors":"Ilia V. Voronov , Vladislav S. Nikolaev , Alexey V. Timofeev , Vladimir V. Stegailov","doi":"10.1016/j.jnucmat.2024.155483","DOIUrl":"10.1016/j.jnucmat.2024.155483","url":null,"abstract":"<div><div>A bcc iron bicrystal in contact with liquid lead is studied in molecular dynamics simulations to describe the atomistic mechanism of liquid metal corrosion in the activation controlled case. In this process, the main structural features involved are Fe grain boundaries and Fe-Pb interfaces. The atomistic model considered reveals that the interplay of atomic processes such as surface self-diffusion of Fe and dissolution of Fe into Pb determines the mechanism and kinetics law of liquid metal corrosion. Analysis of the proposed mechanism explains the dependence between the kinetics of liquid metal corrosion and the grain size of the specimen.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155483"},"PeriodicalIF":2.8,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142587028","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.2024.155480
R.A. Rymzhanov , A.E. Volkov , V.A. Skuratov
Formation of tracks of swift heavy ions decelerating in the electronic stopping regime in CeO2 was studied, combining the Monte Carlo code TREKIS with molecular dynamics. We show that strong lattice disordering (melting) followed by structure recovery form finally a damaged ion track consisting of a discontinuous crystalline region in CeO2. Normal ion impacts result in appearance of spherical crystalline hillocks on CeO2 surface. The solid-vacuum interface strongly suppresses the recrystallization of the near-surface layers, forming conically shaped tracks with several tens of nanometers lengths. Grazing ion irradiation induces intensive material expulsion from the surface forming finally grooves surrounded by nanohillocks. The processes of surface nanostructures formation is similar to those observed previously in CaF2 which has the similar crystalline structure, however requires much longer recrystallization time. Recent experimental data confirm the simulation results.
{"title":"Bulk, overlap and surface effects of swift heavy ions in CeO2","authors":"R.A. Rymzhanov , A.E. Volkov , V.A. Skuratov","doi":"10.1016/j.jnucmat.2024.155480","DOIUrl":"10.1016/j.jnucmat.2024.155480","url":null,"abstract":"<div><div>Formation of tracks of swift heavy ions decelerating in the electronic stopping regime in CeO<sub>2</sub> was studied, combining the Monte Carlo code TREKIS with molecular dynamics. We show that strong lattice disordering (melting) followed by structure recovery form finally a damaged ion track consisting of a discontinuous crystalline region in CeO<sub>2</sub>. Normal ion impacts result in appearance of spherical crystalline hillocks on CeO<sub>2</sub> surface. The solid-vacuum interface strongly suppresses the recrystallization of the near-surface layers, forming conically shaped tracks with several tens of nanometers lengths. Grazing ion irradiation induces intensive material expulsion from the surface forming finally grooves surrounded by nanohillocks. The processes of surface nanostructures formation is similar to those observed previously in CaF<sub>2</sub> which has the similar crystalline structure, however requires much longer recrystallization time. Recent experimental data confirm the simulation results.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155480"},"PeriodicalIF":2.8,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571580","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}
The coarsening of helium (He) bubbles in nickel-based alloys significantly impacts their service performance. Understanding the underlying mechanisms is crucial for ensuring the long-term durability and reliability of these alloys in reactor radiation environments. Molecular dynamics simulations of single bubble growth at temperatures of 300 and 900 K were conducted using the sequential He atom injection method to investigate the He bubble growth and evolution in nickel. A noteworthy phenomenon observed during bubble growth is the detachment of punched prismatic loops. The critical bubble size for punched loop detachment can be reduced by growing the bubble at a slower rate or lower temperature. The reduction is attributed to the additional time available for the punched loop to dissociate or the higher pressure within the bubble pushing it out. Meanwhile, the formation mechanism of bubble-loop complexes is explored through the interaction of punched loops with nearby punched loops or bubbles. In addition, the integration of these simulation results with variations in material mechanical performance yields valuable insights for interpreting material degradation. This study provides a foundation for improving in-reactor service performance, contributing to a broader understanding of the complex interplay between helium bubble coarsening and material behavior.
镍基合金中的氦（He）气泡变粗会严重影响其使用性能。了解其基本机制对于确保这些合金在反应堆辐射环境中的长期耐用性和可靠性至关重要。为了研究氦气泡在镍中的生长和演化，我们采用连续注入氦原子的方法，对温度为 300 和 900 K 的单个气泡生长进行了分子动力学模拟。在气泡生长过程中观察到的一个值得注意的现象是冲孔棱柱环的脱离。以更慢的速度或更低的温度生长气泡，可以减小冲孔环脱离的临界气泡尺寸。这种减小可归因于冲孔环有更多的时间解离或气泡内更高的压力将其挤出。同时，通过打孔环与附近打孔环或气泡的相互作用，探索了气泡环复合物的形成机制。此外，将这些模拟结果与材料力学性能的变化相结合，还能为解释材料降解提供有价值的见解。这项研究为提高反应器内的服务性能奠定了基础，有助于更广泛地了解氦气泡粗化与材料行为之间复杂的相互作用。
{"title":"Molecular dynamics simulation of punched loop detachment during helium bubble growth in nickel","authors":"A-Li Wen , He-Fei Huang , Zhen-Bo Zhu , Wei Zhang , Fei-Fei Zhang , Cui-Lan Ren , Ping Huai","doi":"10.1016/j.jnucmat.2024.155479","DOIUrl":"10.1016/j.jnucmat.2024.155479","url":null,"abstract":"<div><div>The coarsening of helium (He) bubbles in nickel-based alloys significantly impacts their service performance. Understanding the underlying mechanisms is crucial for ensuring the long-term durability and reliability of these alloys in reactor radiation environments. Molecular dynamics simulations of single bubble growth at temperatures of 300 and 900 K were conducted using the sequential He atom injection method to investigate the He bubble growth and evolution in nickel. A noteworthy phenomenon observed during bubble growth is the detachment of punched prismatic loops. The critical bubble size for punched loop detachment can be reduced by growing the bubble at a slower rate or lower temperature. The reduction is attributed to the additional time available for the punched loop to dissociate or the higher pressure within the bubble pushing it out. Meanwhile, the formation mechanism of bubble-loop complexes is explored through the interaction of punched loops with nearby punched loops or bubbles. In addition, the integration of these simulation results with variations in material mechanical performance yields valuable insights for interpreting material degradation. This study provides a foundation for improving in-reactor service performance, contributing to a broader understanding of the complex interplay between helium bubble coarsening and material behavior.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155479"},"PeriodicalIF":2.8,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652879","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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