Journal of Nuclear Materials最新文献

英文中文

A new model of fission gas bubble growth and mechanism analysis for U-xZr fuels

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155578

Xiaoxiao Mao , Xingdi Chen , Xiaobin Jian , Feng Yan , Shurong Ding

U-xZr alloys have a promising application prospect in advanced nuclear fuel elements, and their macroscale volume growth under the extreme service environments are attracting more attention. In this study, innovative volume growth modeling and mechanism analysis are performed for various U-xZr alloys under different operation conditions. Specially, based on the creep test results in the references, the macroscale thermal creep models are newly developed for solid U-xZr alloys within a temperature range, implicitly reflecting the effects of phase fraction; for the bubble contained region of equivalent spherical fuel grain, the established thermal creep models are involved in the mechanical constitutive relations for the solid fuel skeleton; the finite element equations are derived for the displacement fields of bubble contained region and numerically implemented, obtaining the multi-level variables of macroscale volume growth, the local porosity and the average porosity. The predictions of irradiation swelling for different U-xZr alloys agree well with the experimental data at 743 K or 903 K; the fast-swelling phenomena due to various thermal creep contributions could be captured, demonstrating the progressiveness of the developed new models and algorithms. The numerical simulation results indicate that: (1) under the irradiation temperature of 603 K or 703 K, dislocation creep mechanism of fuel skeleton is dominated, due to higher internal and external pressure differences; (2) at the high temperatures of 803 K and 903 K, the thermal diffusion creep deformations of fuel skeleton contribute dominantly to the macroscale volume growth of U-xZr alloys over the whole irradiation process; (3) under zero external pressure the sharp increase phenomena of fission gas swelling become more and more distinct with the rise of irradiation temperature, stemming from the quickened diffusion of fission gas atom and the enhanced creep deformations of fuel skeleton; at 903 K the fuel skeleton is prone to creep deformation, leading to significant inhibition of bubble growth by a small external pressure. This research provides important theoretical models and algorithms for simulation of the irradiation-induced thermo-mechanical behaviors in U-xZr-based fuel elements or assemblies.

{"title":"A new model of fission gas bubble growth and mechanism analysis for U-xZr fuels","authors":"Xiaoxiao Mao , Xingdi Chen , Xiaobin Jian , Feng Yan , Shurong Ding","doi":"10.1016/j.jnucmat.2024.155578","DOIUrl":"10.1016/j.jnucmat.2024.155578","url":null,"abstract":"<div><div>U-<em>x</em>Zr alloys have a promising application prospect in advanced nuclear fuel elements, and their macroscale volume growth under the extreme service environments are attracting more attention. In this study, innovative volume growth modeling and mechanism analysis are performed for various U-<em>x</em>Zr alloys under different operation conditions. Specially, based on the creep test results in the references, the macroscale thermal creep models are newly developed for solid U-<em>x</em>Zr alloys within a temperature range, implicitly reflecting the effects of phase fraction; for the bubble contained region of equivalent spherical fuel grain, the established thermal creep models are involved in the mechanical constitutive relations for the solid fuel skeleton; the finite element equations are derived for the displacement fields of bubble contained region and numerically implemented, obtaining the multi-level variables of macroscale volume growth, the local porosity and the average porosity. The predictions of irradiation swelling for different U-<em>x</em>Zr alloys agree well with the experimental data at 743 K or 903 K; the fast-swelling phenomena due to various thermal creep contributions could be captured, demonstrating the progressiveness of the developed new models and algorithms. The numerical simulation results indicate that: (1) under the irradiation temperature of 603 K or 703 K, dislocation creep mechanism of fuel skeleton is dominated, due to higher internal and external pressure differences; (2) at the high temperatures of 803 K and 903 K, the thermal diffusion creep deformations of fuel skeleton contribute dominantly to the macroscale volume growth of U-<em>x</em>Zr alloys over the whole irradiation process; (3) under zero external pressure the sharp increase phenomena of fission gas swelling become more and more distinct with the rise of irradiation temperature, stemming from the quickened diffusion of fission gas atom and the enhanced creep deformations of fuel skeleton; at 903 K the fuel skeleton is prone to creep deformation, leading to significant inhibition of bubble growth by a small external pressure. This research provides important theoretical models and algorithms for simulation of the irradiation-induced thermo-mechanical behaviors in U-<em>x</em>Zr-based fuel elements or assemblies.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155578"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170572","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}

引用次数: 0

Molecular dynamics simulations of interaction between a super edge dislocation and interstitial dislocation loops in irradiated L12-Ni3Al

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155541

Cheng Chen , Dongyang Qin , Yiding Wang , Fei Xu , Jun Song

The study employed MD simulations to investigate the interactions between a

〈 1 \bar{1} 0 〉

super-edge dislocation, consisting of the four Shockley partials, and interstitial dislocation loops (IDLs) in irradiated

L 1_{2}

-Ni₃Al. Accounting for symmetry breakage in the

L 1_{2}

lattice, the superlattice planar faults with four distinct fault vectors have been considered for different IDL configurations. The detailed dislocation reactions and structural evolution events were identified as the four partials interacted with various IDL configurations. The slipping characteristics of Shockley partials within the IDLs and the resultant shearing and looping mechanisms were also clarified, revealing distinct energetic transition states determined by the fault vectors after the Shockley partials sweeping the IDL. Furthermore, significant variations in critical resolved shear stress (CRSS) required for the super-edge dislocation to move past the IDL were observed, attributed to various sizes and faulted vectors of enclosed superlattice planar faults in the IDLs. The current study extends the existing dislocation-IDL interaction theory from pristine FCC to

L 1_{2}

lattice, advances the understanding of irradiation hardening effects in

L 1_{2}

N i_{3}

Al, and suggests potential applicability to other

L 1_{2}

systems.

{"title":"Molecular dynamics simulations of interaction between a super edge dislocation and interstitial dislocation loops in irradiated L12-Ni3Al","authors":"Cheng Chen , Dongyang Qin , Yiding Wang , Fei Xu , Jun Song","doi":"10.1016/j.jnucmat.2024.155541","DOIUrl":"10.1016/j.jnucmat.2024.155541","url":null,"abstract":"<div><div>The study employed MD simulations to investigate the interactions between a <span><math><mrow><mo>〈</mo><mrow><mn>1</mn><mover><mn>1</mn><mo>¯</mo></mover><mn>0</mn></mrow><mo>〉</mo></mrow></math></span> super-edge dislocation, consisting of the four Shockley partials, and interstitial dislocation loops (IDLs) in irradiated <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span>-Ni<sub>3</sub>Al. Accounting for symmetry breakage in the <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span> lattice, the superlattice planar faults with four distinct fault vectors have been considered for different IDL configurations. The detailed dislocation reactions and structural evolution events were identified as the four partials interacted with various IDL configurations. The slipping characteristics of Shockley partials within the IDLs and the resultant shearing and looping mechanisms were also clarified, revealing distinct energetic transition states determined by the fault vectors after the Shockley partials sweeping the IDL. Furthermore, significant variations in critical resolved shear stress (CRSS) required for the super-edge dislocation to move past the IDL were observed, attributed to various sizes and faulted vectors of enclosed superlattice planar faults in the IDLs. The current study extends the existing dislocation-IDL interaction theory from pristine FCC to <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span> lattice, advances the understanding of irradiation hardening effects in <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span>-<span><math><mrow><mi>N</mi><msub><mi>i</mi><mn>3</mn></msub></mrow></math></span>Al, and suggests potential applicability to other <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span> systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155541"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171148","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}

引用次数: 0

Accelerated prediction of lattice thermal conductivity of Zirconium and its alloys: A machine learning potential method

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155603

Fan Yang , Di Wang , Jiaxuan Si , Jianqiao Yu , Zhen Xie , Xiaoyong Wu , Yuexia Wang

Zirconium alloy coating is an important direction for the modification of nuclear cladding materials. Thermal conductivity is a critical property of cladding materials. With extensively studying phonon-electron non-equilibrium energy transfer processes in the thermal transport of zirconium alloy coating, to distinguish the contributions from phonon and electron thermal conductivity of Zr alloys becomes crucial and necessary. In this work, we successfully predicted the lattice thermal conductivities of zirconium, Zr-Sn and Zr-Nb using machine learning potentials. Sn and Nb doping leads to a significant decrease in lattice thermal conductivity, which is mainly due to the alterations in phonon group velocity and phonon scattering. The larger atomic mass of doping elements and weakened interatomic interactions of Zr-Nb together lead to a significant decrease in phonon group velocity. Doping Sn and Nb also increases phonon-phonon scattering rate and three-phonon scattering channels, resulting in a shortening in phonon lifetime and a decrease in lattice thermal conductivity.

{"title":"Accelerated prediction of lattice thermal conductivity of Zirconium and its alloys: A machine learning potential method","authors":"Fan Yang , Di Wang , Jiaxuan Si , Jianqiao Yu , Zhen Xie , Xiaoyong Wu , Yuexia Wang","doi":"10.1016/j.jnucmat.2024.155603","DOIUrl":"10.1016/j.jnucmat.2024.155603","url":null,"abstract":"<div><div>Zirconium alloy coating is an important direction for the modification of nuclear cladding materials. Thermal conductivity is a critical property of cladding materials. With extensively studying phonon-electron non-equilibrium energy transfer processes in the thermal transport of zirconium alloy coating, to distinguish the contributions from phonon and electron thermal conductivity of Zr alloys becomes crucial and necessary. In this work, we successfully predicted the lattice thermal conductivities of zirconium, Zr-Sn and Zr-Nb using machine learning potentials. Sn and Nb doping leads to a significant decrease in lattice thermal conductivity, which is mainly due to the alterations in phonon group velocity and phonon scattering. The larger atomic mass of doping elements and weakened interatomic interactions of Zr-Nb together lead to a significant decrease in phonon group velocity. Doping Sn and Nb also increases phonon-phonon scattering rate and three-phonon scattering channels, resulting in a shortening in phonon lifetime and a decrease in lattice thermal conductivity.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155603"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171594","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}

引用次数: 0

In-situ strain behavior and BISON simulations of Zircaloy cladding subjected to temperature cycling separate-effects tests in a steam environment

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155570

Jennifer I. Espersen , Nathan A. Capps , Mackenzie J. Ridley , Sam B. Bell , Nicholas R. Brown

Understanding fuel system performance during anticipated transients without scram (ATWSs) in boiling water reactors (BWRs) is necessary for refining current and future safety limits. High-fidelity material models and simulations are fundamental to rigorous assessment of zirconium-based cladding performance. However, experimental thermomechanical data during simulated ATWSs to validate these modes are limited. To provide relevant in-situ data, Zircaloy-4 cladding was subjected to cyclic heating in a steam environment to simulate an out-of-pile BWR ATWS. Digital image correlation was used to capture the cladding strain behavior in-situ for comparison against simulations using the BISON finite element code. Conventional high-temperature models were compared using multiple schemes to gain a better understanding of the applicability of three BISON models to BWR ATWS: (1) the default combination of creep models in BISON, (2) the high-temperature Erbacher model alone, and (3) the low-temperature Limback-Andersson model alone. The cases run with the Limback-Andersson model alone produced the lowest root mean square error (RMSE). The lowest RMSE for the Limback-Andersson model alone was 0.659%, and the highest RMSE reported was 4.22%. A data gap within the model in the temperature regime of interest was also identified, and to account for this gap, the current model in BISON is linearly interpolated between two separate datasets. This evaluation highlights the need to either develop a new model or to improve the existing model to capture transient creep effects resulting from a cyclic temperature transient.

{"title":"In-situ strain behavior and BISON simulations of Zircaloy cladding subjected to temperature cycling separate-effects tests in a steam environment","authors":"Jennifer I. Espersen , Nathan A. Capps , Mackenzie J. Ridley , Sam B. Bell , Nicholas R. Brown","doi":"10.1016/j.jnucmat.2024.155570","DOIUrl":"10.1016/j.jnucmat.2024.155570","url":null,"abstract":"<div><div>Understanding fuel system performance during anticipated transients without scram (ATWSs) in boiling water reactors (BWRs) is necessary for refining current and future safety limits. High-fidelity material models and simulations are fundamental to rigorous assessment of zirconium-based cladding performance. However, experimental thermomechanical data during simulated ATWSs to validate these modes are limited. To provide relevant in-situ data, Zircaloy-4 cladding was subjected to cyclic heating in a steam environment to simulate an out-of-pile BWR ATWS. Digital image correlation was used to capture the cladding strain behavior in-situ for comparison against simulations using the BISON finite element code. Conventional high-temperature models were compared using multiple schemes to gain a better understanding of the applicability of three BISON models to BWR ATWS: (1) the default combination of creep models in BISON, (2) the high-temperature Erbacher model alone, and (3) the low-temperature Limback-Andersson model alone. The cases run with the Limback-Andersson model alone produced the lowest root mean square error (RMSE). The lowest RMSE for the Limback-Andersson model alone was 0.659%, and the highest RMSE reported was 4.22%. A data gap within the model in the temperature regime of interest was also identified, and to account for this gap, the current model in BISON is linearly interpolated between two separate datasets. This evaluation highlights the need to either develop a new model or to improve the existing model to capture transient creep effects resulting from a cyclic temperature transient.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155570"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170555","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}

引用次数: 0

Influence of transmutation products on the thermophysical properties of eutectic NaCl-UCl3 fuel salt in a fast-spectrum molten salt reactor

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155572

Sudipta Paul , Siamak Attarian , Massimiliano Fratoni , Dane Morgan , Izabela Szlufarska

During the operation of fast-spectrum liquid-fueled molten salt reactors (MSRs), the salt composition of the fuel salt changes due to the generation of transmutation products (TPs) with the depletion of fuel, which can influence the thermophysical properties of the salt. Here, we evaluate changes in thermophysical properties of NaCl-UCl₃, such as density, viscosity, heat capacity, and thermal conductivity, due to generation of TPs after a representative burn-up of 180 MWd/Kg-HM. Concentration of TPs was determined from Monte Carlo simulations. Thermophysical properties were evaluated using semi-empirical models with input from experiments, ab initio molecular dynamics, and machine learning potentials. Our analysis predicts that although the aforementioned properties of the salt mixture are altered after the burn-up period, the concentration of TPs is small enough so that the overall changes in these properties in the fuel salt are likely not significant for nuclear applications.

{"title":"Influence of transmutation products on the thermophysical properties of eutectic NaCl-UCl3 fuel salt in a fast-spectrum molten salt reactor","authors":"Sudipta Paul , Siamak Attarian , Massimiliano Fratoni , Dane Morgan , Izabela Szlufarska","doi":"10.1016/j.jnucmat.2024.155572","DOIUrl":"10.1016/j.jnucmat.2024.155572","url":null,"abstract":"<div><div>During the operation of fast-spectrum liquid-fueled molten salt reactors (MSRs), the salt composition of the fuel salt changes due to the generation of transmutation products (TPs) with the depletion of fuel, which can influence the thermophysical properties of the salt. Here, we evaluate changes in thermophysical properties of NaCl-UCl<sub>3</sub>, such as density, viscosity, heat capacity, and thermal conductivity, due to generation of TPs after a representative burn-up of 180 MWd/Kg-HM. Concentration of TPs was determined from Monte Carlo simulations. Thermophysical properties were evaluated using semi-empirical models with input from experiments, <em>ab initio</em> molecular dynamics, and machine learning potentials. Our analysis predicts that although the aforementioned properties of the salt mixture are altered after the burn-up period, the concentration of TPs is small enough so that the overall changes in these properties in the fuel salt are likely not significant for nuclear applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155572"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170560","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}

引用次数: 0

Homogeneous precipitation of thorium oxalate: Structural, kinetic, and morphological aspects

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155574

A. Zakharanka , L. Gubbels , B. Acevedo , M. Verwerft , V. Tyrpekl

Thorium oxalate hexahydrate, Th(C₂O₄)₂·6H₂O was produced by homogeneous precipitation from a thorium nitrate solution through the decomposition of oxamic acid (NH₂COCOOH). Typically, lanthanide oxalates and actinide oxalates are prepared by heterogeneous precipitation using oxalic acid (COOH)₂. However, in the present homogeneous precipitation reaction, oxalic acid was slowly generated by the acid-catalyzed hydrolysis of oxamic acid. While heterogeneous precipitation is rapid and typically yields small microcrystals, the slow generation of the precipitation agent (oxalic acid) during homogeneous precipitation resulted in the formation of large thorium oxalate crystals with atypical morphology. The reaction exhibited first-order kinetics and was assessed at 443, 453 and 463 K (70, 80, 90 °C). The morphology of crystals obtained at these different temperatures were investigated. Additionally, the sensitivity of thorium oxalate hexahydrate to drying conditions and its decomposition during calcination to ThO₂ were examined. Thorium oxalate hexahydrate tends to lose crystalline water, resulting in transition phases toward the dihydrate when dried under vacuum at 313 K (40 °C). This loss of crystalline water was not observed when drying was performed under ambient conditions. The further decomposition of the oxalate dihydrate to ThO₂ followed the well-known decomposition path. The developed reaction is affordable, convenient, and does not require demanding apparatus, making it a versatile preparation route for various thorium oxalate crystals of variable morphology suitable for crystallographic studies or applications demanding powders with large particle sizes.

{"title":"Homogeneous precipitation of thorium oxalate: Structural, kinetic, and morphological aspects","authors":"A. Zakharanka , L. Gubbels , B. Acevedo , M. Verwerft , V. Tyrpekl","doi":"10.1016/j.jnucmat.2024.155574","DOIUrl":"10.1016/j.jnucmat.2024.155574","url":null,"abstract":"<div><div>Thorium oxalate hexahydrate, Th(C<sub>2</sub>O<sub>4</sub>)<sub>2</sub>·6H<sub>2</sub>O was produced by homogeneous precipitation from a thorium nitrate solution through the decomposition of oxamic acid (NH<sub>2</sub>COCOOH). Typically, lanthanide oxalates and actinide oxalates are prepared by heterogeneous precipitation using oxalic acid (COOH)<sub>2</sub>. However, in the present homogeneous precipitation reaction, oxalic acid was slowly generated by the acid-catalyzed hydrolysis of oxamic acid. While heterogeneous precipitation is rapid and typically yields small microcrystals, the slow generation of the precipitation agent (oxalic acid) during homogeneous precipitation resulted in the formation of large thorium oxalate crystals with atypical morphology. The reaction exhibited first-order kinetics and was assessed at 443, 453 and 463 K (70, 80, 90 °C). The morphology of crystals obtained at these different temperatures were investigated. Additionally, the sensitivity of thorium oxalate hexahydrate to drying conditions and its decomposition during calcination to ThO<sub>2</sub> were examined. Thorium oxalate hexahydrate tends to lose crystalline water, resulting in transition phases toward the dihydrate when dried under vacuum at 313 K (40 °C). This loss of crystalline water was not observed when drying was performed under ambient conditions. The further decomposition of the oxalate dihydrate to ThO<sub>2</sub> followed the well-known decomposition path. The developed reaction is affordable, convenient, and does not require demanding apparatus, making it a versatile preparation route for various thorium oxalate crystals of variable morphology suitable for crystallographic studies or applications demanding powders with large particle sizes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155574"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170562","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}

引用次数: 0

The effect of proton irradiation on dealloying of Alloy 800 in an aqueous environment

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155554

M. Rezvanian, H. Gholamzadeh, K. Daub, F. Long, M.R. Daymond, S.Y. Persaud

The effect of prior proton irradiation on the subsequent corrosion and dealloying of Alloy 800 was investigated in a 50 wt. % NaOH caustic solution at 140 °C. Dealloying of irradiated (1 dpa) and non-irradiated Alloy 800 was characterized at the nanoscale to provide mechanistic insight. A porous surface film enriched in Ni was formed on both irradiated and non-irradiated materials due to selective dissolution of Fe and Cr. However, the microstructure and chemistry in the dealloyed layer was altered by irradiation, including deeper dealloyed layer penetration, finer porosity, and changes in the classical core-shell ligament structure in the irradiated material.

引用次数: 0

Molecular dynamics simulation of retention and bubble formation in tungsten with carbon impurities under high flux deuterium irradiation

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155576

Yuan Xiang, Liqun Shi, Bin Zhang

W is a critical plasma-facing material in fusion reactors. However, extreme conditions subject W to high-flux D plasma irradiation, resulting in D retention and bubble formation. Importantly, C impurities in the plasma can have a complex effect on D behavior in the W. Due to experimental limitations, the impact of C on the evolution of irradiated W and the underlying mechanisms remains poorly understood. In this study, we employ molecular dynamics simulations to investigate the effects of C on the retention of D in W. The introduction forms of C include irradiation and direct doping. First, Irradiation can cause C to be retained, at the same time, the structure of W is severely damaged. W atoms in the subsurface are bombarded and displaced to form interstitial atoms and even be sputtered out of the W matrix. Statistical results show that the sputtering of W is primarily determined by the energy of impact C, with temperature having relatively minor effects. C irradiation forms a W-C mixed layer on the W surface. This layer has a disordered structure and is mixed with self-interstitial atoms and C atoms. D atoms incident from above are easily intercepted and captured by this layer and are difficult to migrate deeper. Second, C randomly doped in W significantly promotes the retention of D. The higher the C content, the higher the retention rate. Furthermore, the supersaturated accumulation of retained D leads to the formation of D bubbles. D undergoes an evolutionary process of forming molecules, clusters, and finally D bubbles. This work provides new insights into the W-C-D interaction mechanisms in fusion reactors, offering important theoretical support for the selection of W as a plasma-facing material.

{"title":"Molecular dynamics simulation of retention and bubble formation in tungsten with carbon impurities under high flux deuterium irradiation","authors":"Yuan Xiang, Liqun Shi, Bin Zhang","doi":"10.1016/j.jnucmat.2024.155576","DOIUrl":"10.1016/j.jnucmat.2024.155576","url":null,"abstract":"<div><div>W is a critical plasma-facing material in fusion reactors. However, extreme conditions subject W to high-flux D plasma irradiation, resulting in D retention and bubble formation. Importantly, C impurities in the plasma can have a complex effect on D behavior in the W. Due to experimental limitations, the impact of C on the evolution of irradiated W and the underlying mechanisms remains poorly understood. In this study, we employ molecular dynamics simulations to investigate the effects of C on the retention of D in W. The introduction forms of C include irradiation and direct doping. First, Irradiation can cause C to be retained, at the same time, the structure of W is severely damaged. W atoms in the subsurface are bombarded and displaced to form interstitial atoms and even be sputtered out of the W matrix. Statistical results show that the sputtering of W is primarily determined by the energy of impact C, with temperature having relatively minor effects. C irradiation forms a W-C mixed layer on the W surface. This layer has a disordered structure and is mixed with self-interstitial atoms and C atoms. D atoms incident from above are easily intercepted and captured by this layer and are difficult to migrate deeper. Second, C randomly doped in W significantly promotes the retention of D. The higher the C content, the higher the retention rate. Furthermore, the supersaturated accumulation of retained D leads to the formation of D bubbles. D undergoes an evolutionary process of forming molecules, clusters, and finally D bubbles. This work provides new insights into the W-C-D interaction mechanisms in fusion reactors, offering important theoretical support for the selection of W as a plasma-facing material.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155576"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171600","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}

引用次数: 0

Experimental investigation and theoretical validation of failure mechanism caused by tellurium in Inconel 718

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155670

Guoying Li , Zhanqiang Liu , Bing Wang , Zhao Qian , Zongde Kou , Yingwen Li

Tellurium (Te) in molten salt reactors (MSR) induces intergranular corrosion and material failure in nickel-based superalloys. This paper investigates the failure mechanism using experiments and first-principles calculations. Four characteristics of tellurides are experimentally identified, revealing the causes of Te-induced failure. The effects of Te doping on segregation energy, grain boundary binding energy, charge density and density of states are analyzed. The superlattice heterojunction interface model is developed to compute the interfacial adhesion work. A cross-scale transfer mechanism connecting electronic interactions, atomic behavior, microstructure and Te-induced failure is established to explain the degradation of mechanical properties.

引用次数: 0

Release of ion-implanted 3He and D from tungsten under subsequent 4He ion irradiation

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155612

A. Umerenkova, Z. Harutyunyan, O.V. Ogorodnikova, Y. Gasparyan, N. Ostojic, V. Efimov

The efficiency of deuterium (D) and ³He atoms removal from polycrystalline tungsten (W) by irradiation with 3 keV ⁴He ions at room temperature was investigated. It was shown that up to 90% of He isotopes are replaced by post-irradiation with another He isotopes in both directions when saturation with He atoms in W is achieved. There are at least four processes of He isotope exchange and D-He exchange, namely, (1) collision, (2) sputtering, (3) bubble bursting and (4) replacement of previously trapped atom by post-implanted atom. It was shown that the sputtering and bubble bursting play a minor role in the removal of D and He by 3 keV ⁴He ions in comparison with collision and replacement processes. In the case of He isotope exchange, the dominate process is most likely a replacement. Replacement occurs even at room temperature regardless on the binding energy of He with defects, probably, by decreasing of the binding energy of He or D atoms with defects when He atoms during sequential irradiation approach them. In the case of D-He exchange, both collision cascade and replacement play a main role in the D removal by 3 keV ⁴He ions. The mechanism of the D and He removal by He ions is still need further study.

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