Journal of Nuclear Materials最新文献

Unravelling the potential of iodine isotopic exchange in CH3131I capture by K127I-impregnated activated carbons

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

Journal of Nuclear Materials

Pub Date : 2025-03-01 DOI: 10.1016/j.jnucmat.2025.155719

K. Abbas , M. Chebbi , B. Azambre , C. Monsanglant-Louvet , B. Marcillaud , A. Roynette

The efficient capture of radioactive methyl iodide (CH₃¹³¹I) is a critical issue for nuclear safety and radioprotection. Co-impregnated activated carbons (AC), with triethylenediamine (TEDA) and potassium iodide (K¹²⁷I), are widely employed for this purpose. However, the specific role of KI in CH₃¹³¹I retention through isotopic exchange reaction remains poorly understood. This study provides groundbreaking insights by systematically investigating the retention behavior of KI/AC versus non-impregnated activated carbons (NI AC) under different operating conditions. Advanced characterization techniques, including N₂ porosimetry, high-resolution transmission electron microscopy (HRTEM), and H₂O adsorption isotherms, were first employed to elucidate the structural and chemical properties of the adsorbents. Subsequently, CH₃¹³¹I retention tests were conducted by measuring the Decontamination Factors (DF) at various configurations covering a broad range of relative humidities (RH) (20–90 %), temperatures (20–96 °C), residence times (0.125–0.5 s) and elution times (1–18 h). Results reveal that while NI AC exhibits a drastic performance decline at high RH attributable to water physisorption, KI/AC demonstrates enhanced retention, counterbalancing moisture effects via isotopic exchange. Furthermore, elevated temperatures significantly amplify DF for KI/AC, unveiling for the first time the thermally activated nature of the isotopic exchange mechanism. Prolonged residence time further enhance performance for KI/AC compared to NI AC, suggesting multiple mechanistic steps in isotopic exchange reaction. Consequently, a detailed mechanism for this reaction has been proposed.

This work advances the understanding of CH₃¹³¹I capture mechanisms ensuring improved performance under diverse nuclear safety scenarios.

{"title":"Unravelling the potential of iodine isotopic exchange in CH3131I capture by K127I-impregnated activated carbons","authors":"K. Abbas , M. Chebbi , B. Azambre , C. Monsanglant-Louvet , B. Marcillaud , A. Roynette","doi":"10.1016/j.jnucmat.2025.155719","DOIUrl":"10.1016/j.jnucmat.2025.155719","url":null,"abstract":"<div><div>The efficient capture of radioactive methyl iodide (CH<sub>3</sub><sup>131</sup>I) is a critical issue for nuclear safety and radioprotection. Co-impregnated activated carbons (AC), with triethylenediamine (TEDA) and potassium iodide (K<sup>127</sup>I), are widely employed for this purpose. However, the specific role of KI in CH<sub>3</sub><sup>131</sup>I retention through isotopic exchange reaction remains poorly understood. This study provides groundbreaking insights by systematically investigating the retention behavior of KI/AC <em>versus</em> non-impregnated activated carbons (NI AC) under different operating conditions. Advanced characterization techniques, including N<sub>2</sub> porosimetry, high-resolution transmission electron microscopy (HRTEM), and H<sub>2</sub>O adsorption isotherms, were first employed to elucidate the structural and chemical properties of the adsorbents. Subsequently, CH<sub>3</sub><sup>131</sup>I retention tests were conducted by measuring the Decontamination Factors (DF) at various configurations covering a broad range of relative humidities (RH) (20–90 %), temperatures (20–96 °C), residence times (0.125–0.5 s) and elution times (1–18 h). Results reveal that while NI AC exhibits a drastic performance decline at high RH attributable to water physisorption, KI/AC demonstrates enhanced retention, counterbalancing moisture effects <em>via</em> isotopic exchange. Furthermore, elevated temperatures significantly amplify DF for KI/AC, unveiling for the first time the thermally activated nature of the isotopic exchange mechanism. Prolonged residence time further enhance performance for KI/AC compared to NI AC, suggesting multiple mechanistic steps in isotopic exchange reaction. Consequently, a detailed mechanism for this reaction has been proposed.</div><div>This work advances the understanding of CH<sub>3</sub><sup>131</sup>I capture mechanisms ensuring improved performance under diverse nuclear safety scenarios.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155719"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548946","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

Kr10+ irradiation stability and strain accumulation of MgONd2(Zr1-xCex)2O7 composite ceramics for inert matrix fuel

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

Journal of Nuclear Materials

Pub Date : 2025-03-01 DOI: 10.1016/j.jnucmat.2025.155722

Yijie Tang , Jin Wang , Junxia Wang , Long Kang , Tongming Zhang , Jun Li , Yan Wang , Xusheng Li , Yanping Yang

In this work, MgONd₂(Zr_1-_xCe_x)₂O₇ (M-NZC_x) composite ceramics, a potential matrix, used for inert matrix fuel were subjected to 3.5 MeV Kr¹⁰⁺ irradiation with fluences of 1 × 10¹⁴–5 × 10¹⁵ ions/cm². The phase transition from pyrochlore (P) to defect fluorite (F) structures (P-F transition) in NZC_x phase, as well as the lattice expansion, amorphization and strain accumulation in both MgO and pyrochlore phases in post-irradiated M-NZC_x composite ceramics were systematically investigated. Specifically, the lattice expansion ratios (Rs) of different phases in M-NZC_x samples irradiated with the same fluence of Kr¹⁰⁺ ions were ranged as follows, Rs_M < Rs_P < Rs_F. The further research on irradiation response of different phases in typical M-NZC_0.3 sample indicated that MgO exhibited superior irradiation stability to pyrochlore phase. Additionally, the strain accumulation induced by Kr¹⁰⁺ bombardment was observed generally in both MgO and NZC_0.3 pyrochlore phases, and the defect-induced strain in MgO crystalline was more pronounced than NZC_0.3 pyrochlore phase. Interestingly, the strain accumulation resulting from Kr¹⁰⁺ irradiation was observed to be preferentially oriented along directions with lower atomic density. This study might provide a new perspective for understanding the irradiation stability of MgONd₂Zr₂O₇ based composite ceramics in elastic collision cascade by krypton ions.

{"title":"Kr10+ irradiation stability and strain accumulation of MgONd2(Zr1-xCex)2O7 composite ceramics for inert matrix fuel","authors":"Yijie Tang , Jin Wang , Junxia Wang , Long Kang , Tongming Zhang , Jun Li , Yan Wang , Xusheng Li , Yanping Yang","doi":"10.1016/j.jnucmat.2025.155722","DOIUrl":"10.1016/j.jnucmat.2025.155722","url":null,"abstract":"<div><div>In this work, MgO<img>Nd<sub>2</sub>(Zr<sub>1-</sub><em><sub>x</sub></em>Ce<em><sub>x</sub></em>)<sub>2</sub>O<sub>7</sub> (M-NZC<em><sub>x</sub></em>) composite ceramics, a potential matrix, used for inert matrix fuel were subjected to 3.5 MeV Kr<sup>10+</sup> irradiation with fluences of 1 × 10<sup>14</sup>–5 × 10<sup>15</sup> ions/cm<sup>2</sup>. The phase transition from pyrochlore (P) to defect fluorite (F) structures (P-F transition) in NZC<em><sub>x</sub></em> phase, as well as the lattice expansion, amorphization and strain accumulation in both MgO and pyrochlore phases in post-irradiated M-NZC<em><sub>x</sub></em> composite ceramics were systematically investigated. Specifically, the lattice expansion ratios (<em>Rs</em>) of different phases in M-NZC<em><sub>x</sub></em> samples irradiated with the same fluence of Kr<sup>10+</sup> ions were ranged as follows, <em>Rs</em><sub>M</sub> < <em>Rs</em><sub>P</sub> < <em>Rs</em><sub>F</sub>. The further research on irradiation response of different phases in typical M-NZC<sub>0.3</sub> sample indicated that MgO exhibited superior irradiation stability to pyrochlore phase. Additionally, the strain accumulation induced by Kr<sup>10+</sup> bombardment was observed generally in both MgO and NZC<sub>0.3</sub> pyrochlore phases, and the defect-induced strain in MgO crystalline was more pronounced than NZC<sub>0.3</sub> pyrochlore phase. Interestingly, the strain accumulation resulting from Kr<sup>10+</sup> irradiation was observed to be preferentially oriented along directions with lower atomic density. This study might provide a new perspective for understanding the irradiation stability of MgO<img>Nd<sub>2</sub>Zr<sub>2</sub>O<sub>7</sub> based composite ceramics in elastic collision cascade by krypton ions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155722"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548944","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

Microstructural evolution of neutron irradiated ultrafine-grained austenitic stainless steel

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

Journal of Nuclear Materials

Pub Date : 2025-03-01 DOI: 10.1016/j.jnucmat.2025.155710

Frederic Habiyaremye , Solène Rouland , Bertrand Radiguet , Fabien Cuvilly , Benjamin Klaes , Benoit Tanguy , Joël Malaplate , Christophe Domain , Diogo Goncalves , Marina M. Abramova , Nariman A. Enikeev , Xavier Sauvage , Auriane Etienne

Austenitic stainless steels utilized in-core components of pressurized water reactors are prone to radiation-induced segregation, which leads to the degradation of microstructure and mechanical properties. To improve irradiation resistance, one possible solution is to increase the number density of point defect sinks, such as grain boundaries. For this purpose, ultrafine-grained or nanostructured microstructures are recommended due to their high density of grain boundaries. This paper investigates the microstructural changes in ultrafine-grained 316 austenitic stainless steel exposed to neutron radiation up to 3.9 dpa in irradiation conditions representative of light water reactors. The microstructure at different length scales was analyzed using electron backscattered diffraction, transmission electron microscopy, and atom probe tomography before and after neutron irradiation. The study compares its findings with those of existing literature on coarse-grained austenitic stainless steels to evaluate the benefit of ultrafine-grained 316 austenitic stainless steels regarding irradiation ageing in representative conditions of light water reactors.

{"title":"Microstructural evolution of neutron irradiated ultrafine-grained austenitic stainless steel","authors":"Frederic Habiyaremye , Solène Rouland , Bertrand Radiguet , Fabien Cuvilly , Benjamin Klaes , Benoit Tanguy , Joël Malaplate , Christophe Domain , Diogo Goncalves , Marina M. Abramova , Nariman A. Enikeev , Xavier Sauvage , Auriane Etienne","doi":"10.1016/j.jnucmat.2025.155710","DOIUrl":"10.1016/j.jnucmat.2025.155710","url":null,"abstract":"<div><div>Austenitic stainless steels utilized in-core components of pressurized water reactors are prone to radiation-induced segregation, which leads to the degradation of microstructure and mechanical properties. To improve irradiation resistance, one possible solution is to increase the number density of point defect sinks, such as grain boundaries. For this purpose, ultrafine-grained or nanostructured microstructures are recommended due to their high density of grain boundaries. This paper investigates the microstructural changes in ultrafine-grained 316 austenitic stainless steel exposed to neutron radiation up to 3.9 dpa in irradiation conditions representative of light water reactors. The microstructure at different length scales was analyzed using electron backscattered diffraction, transmission electron microscopy, and atom probe tomography before and after neutron irradiation. The study compares its findings with those of existing literature on coarse-grained austenitic stainless steels to evaluate the benefit of ultrafine-grained 316 austenitic stainless steels regarding irradiation ageing in representative conditions of light water reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155710"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512334","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}

引用次数: 0

Modeling and analysis for the anisotropic irradiation swelling of porous SiC/SiC composites

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

Journal of Nuclear Materials

Pub Date : 2025-03-01 DOI: 10.1016/j.jnucmat.2025.155711

Luning Chen, Jing Zhang, Shurong Ding

SiC/SiC composites are one of the promising engineering materials for nuclear applications. Anisotropic swelling deformations were observed in these materials during irradiation, and the underlying mechanisms should be deeply understood. In this study, a numerical simulation method is developed to predict the irradiation-induced deformations of the as-fabricated SiC/SiC composites. An emphasis is given to the generation of an RVE (Representative Volume Element) model with a pre-existing pore and the assumed residual stress field. Besides, the thermo-mechanical constitutive relations and stress update algorithms for the solid skeleton of porous SiC/SiC composites are developed with their irradiation effects considered comprehensively. Based on the homogenization theory, the calculation models to obtain the macroscopic swelling strains of porous SiC/SiC composites are developed. The good agreements between the predictions and the post-irradiation data of anisotropic swelling validate the effectiveness of the developed models and simulation methods. Research findings indicate that the irradiation creep deformations due to the existing residual stresses and high transient creep rate coefficients lead to the through-thickness size shrinkage of the pre-existing pores, which possibly becomes the dominant mechanism of the negative linear swelling of the SiC/SiC sample during the initial irradiation stage. The effects of the initial residual stress fields and the elastic constitutive relations on the anisotropic irradiation swelling behaviors are investigated. This study lays a foundation for the advanced manufacture of the SiC/SiC composites and the based multi-layer cladding tubes.

{"title":"Modeling and analysis for the anisotropic irradiation swelling of porous SiC/SiC composites","authors":"Luning Chen, Jing Zhang, Shurong Ding","doi":"10.1016/j.jnucmat.2025.155711","DOIUrl":"10.1016/j.jnucmat.2025.155711","url":null,"abstract":"<div><div>SiC/SiC composites are one of the promising engineering materials for nuclear applications. Anisotropic swelling deformations were observed in these materials during irradiation, and the underlying mechanisms should be deeply understood. In this study, a numerical simulation method is developed to predict the irradiation-induced deformations of the as-fabricated SiC/SiC composites. An emphasis is given to the generation of an RVE (Representative Volume Element) model with a pre-existing pore and the assumed residual stress field. Besides, the thermo-mechanical constitutive relations and stress update algorithms for the solid skeleton of porous SiC/SiC composites are developed with their irradiation effects considered comprehensively. Based on the homogenization theory, the calculation models to obtain the macroscopic swelling strains of porous SiC/SiC composites are developed. The good agreements between the predictions and the post-irradiation data of anisotropic swelling validate the effectiveness of the developed models and simulation methods. Research findings indicate that the irradiation creep deformations due to the existing residual stresses and high transient creep rate coefficients lead to the through-thickness size shrinkage of the pre-existing pores, which possibly becomes the dominant mechanism of the negative linear swelling of the SiC/SiC sample during the initial irradiation stage. The effects of the initial residual stress fields and the elastic constitutive relations on the anisotropic irradiation swelling behaviors are investigated. This study lays a foundation for the advanced manufacture of the SiC/SiC composites and the based multi-layer cladding tubes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155711"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512330","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

Structure of the fuel-cladding chemical interaction (FCCI) layer of a high burnup Zr-1Nb nuclear fuel cladding

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

Journal of Nuclear Materials

Pub Date : 2025-03-01 DOI: 10.1016/j.jnucmat.2025.155699

Wulin Song , Xue Han , Huanlin Cheng, Qi Tang, Huacai Wang, Songtao Ji

A comprehensive characterization of the Fuel Cladding Chemical Interaction (FCCI) layer in a Zr-1Nb alloy with a burnup of 41GWd·tU⁻¹ has been performed utilizing primary techniques including Optical Microscopy (OM), Transmission Electron Microscopy (TEM), and Transmission Kikuchi Diffraction (TKD). The results indicate that the FCCI layer characterized in this study is mainly composed of tetragonal zirconia on both the cladding and fuel sides, with monoclinic zirconia in between. Additionally, the strong fiber texture in monoclinic and tetragonal zirconia aligns well with that in the water-side oxide film. The orientation relationships between α-Zr, monoclinic zirconia and tetragonal zirconia are (

\bar{1}

011) _α-Zr || (010) _m-ZrO2 and (10

\bar{1}

)_m-ZrO2 || (100) _t-ZrO2. It appears that the dominant force for texture development in the FCCI formed on this alloy is the α-Zr to m-ZrO₂ and m-ZrO₂ to t-ZrO₂ transformation stress which is independent with metal substrate orientation.

{"title":"Structure of the fuel-cladding chemical interaction (FCCI) layer of a high burnup Zr-1Nb nuclear fuel cladding","authors":"Wulin Song , Xue Han , Huanlin Cheng, Qi Tang, Huacai Wang, Songtao Ji","doi":"10.1016/j.jnucmat.2025.155699","DOIUrl":"10.1016/j.jnucmat.2025.155699","url":null,"abstract":"<div><div>A comprehensive characterization of the Fuel Cladding Chemical Interaction (FCCI) layer in a Zr-1Nb alloy with a burnup of 41GWd·tU<sup>−1</sup> has been performed utilizing primary techniques including Optical Microscopy (OM), Transmission Electron Microscopy (TEM), and Transmission Kikuchi Diffraction (TKD). The results indicate that the FCCI layer characterized in this study is mainly composed of tetragonal zirconia on both the cladding and fuel sides, with monoclinic zirconia in between. Additionally, the strong fiber texture in monoclinic and tetragonal zirconia aligns well with that in the water-side oxide film. The orientation relationships between α-Zr, monoclinic zirconia and tetragonal zirconia are (<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>011) <sub>α-Zr</sub> || (010) <sub>m-ZrO2</sub> and (10<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>)<sub>m-ZrO2</sub> || (100) <sub>t-ZrO2</sub>. It appears that the dominant force for texture development in the FCCI formed on this alloy is the α-Zr to m-ZrO<sub>2</sub> and m-ZrO<sub>2</sub> to t-ZrO<sub>2</sub> transformation stress which is independent with metal substrate orientation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155699"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512335","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

First Post Irradiation Examinations on a fast reactor grade MOX fuel (U0.6,Pu0.4)O2 for Pu-burning application, irradiated in the High Flux Reactor

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

Journal of Nuclear Materials

Pub Date : 2025-03-01 DOI: 10.1016/j.jnucmat.2025.155707

S. van Til , A.V. Fedorov , F. Nindiyasari , F. Charpin-Jacobs , G. Uitslag , F. Pasti , E. D'Agata , N. Chauvin

To explore fuel operational behaviour and material property evolution under Pu-burning conditions for fast reactor application, several (U,Pu)O₂ MOX fuel pins with increased Pu contents (40 %HM) were irradiated in the HFR Petten in the TRABANT-2 experiment. Fuel pin number 2 (pin 2/2 in short) was designed and produced in the CAPRA programme [1], containing annular (U,Pu)O₂ MOX pellets with a Pu content of 40 % (HM), that were fabricated via classic powder metallurgy, loaded into an austenitic steel cladding tube (15–15Ti). The pin was assembled and immersed in a sodium-filled experimental capsule and irradiated in the High Flux Reactor at a linear heat rate (LHR) of 450–480W/cm and with cladding temperatures not exceeding 600 °C. The irradiation was stopped after three irradiation cycles (74 days) after strong mobility of the central hole was observed in the pellets in neutron radiographs, indicating unexpected high central temperatures.

The post-irradiation neutronics analysis, using neutron fluence detectors located close to the pin confirms a maximum LHR of 447 W/cm. Asymmetric central hole growth and relocation was observed in fuel pin regions exceeding LHR of 407 W/cm.

The temperature history was reconstructed, using instrumentation in the HFR and the sample holder and Post Irradiation Examinations (PIE) on this fuel pin are carried out NRG's Hot Cell Laboratories within the European H2020 project PuMMA [2].

This paper presents a reconstruction of the irradiation history, results of a set of non-destructive examinations (NDE) and fission gas release analysis. The underlying phenomenological explanation on the observed asymmetries is presented and preliminary confirmed by a 2D thermal-mechanical model.

{"title":"First Post Irradiation Examinations on a fast reactor grade MOX fuel (U0.6,Pu0.4)O2 for Pu-burning application, irradiated in the High Flux Reactor","authors":"S. van Til , A.V. Fedorov , F. Nindiyasari , F. Charpin-Jacobs , G. Uitslag , F. Pasti , E. D'Agata , N. Chauvin","doi":"10.1016/j.jnucmat.2025.155707","DOIUrl":"10.1016/j.jnucmat.2025.155707","url":null,"abstract":"<div><div>To explore fuel operational behaviour and material property evolution under Pu-burning conditions for fast reactor application, several (U,Pu)O<sub>2</sub> MOX fuel pins with increased Pu contents (40 %HM) were irradiated in the HFR Petten in the TRABANT-2 experiment. Fuel pin number 2 (pin 2/2 in short) was designed and produced in the CAPRA programme [1], containing annular (U,Pu)O<sub>2</sub> MOX pellets with a Pu content of 40 % (HM), that were fabricated via classic powder metallurgy, loaded into an austenitic steel cladding tube (15–15Ti). The pin was assembled and immersed in a sodium-filled experimental capsule and irradiated in the High Flux Reactor at a linear heat rate (LHR) of 450–480W/cm and with cladding temperatures not exceeding 600 °C. The irradiation was stopped after three irradiation cycles (74 days) after strong mobility of the central hole was observed in the pellets in neutron radiographs, indicating unexpected high central temperatures.</div><div>The post-irradiation neutronics analysis, using neutron fluence detectors located close to the pin confirms a maximum LHR of 447 W/cm. Asymmetric central hole growth and relocation was observed in fuel pin regions exceeding LHR of 407 W/cm.</div><div>The temperature history was reconstructed, using instrumentation in the HFR and the sample holder and Post Irradiation Examinations (PIE) on this fuel pin are carried out NRG's Hot Cell Laboratories within the European H2020 project PuMMA [2].</div><div>This paper presents a reconstruction of the irradiation history, results of a set of non-destructive examinations (NDE) and fission gas release analysis. The underlying phenomenological explanation on the observed asymmetries is presented and preliminary confirmed by a 2D thermal-mechanical model.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155707"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548943","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 dose rate on the evolution of microstructure in Zr alloys: A synchrotron microbeam X-ray, TEM, and APT study

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

Journal of Nuclear Materials

Pub Date : 2025-02-28 DOI: 10.1016/j.jnucmat.2025.155721

Ö. Koç , R. Thomas , B. Jenkins , C. Hofer , Z. Hegedüs , U. Lienert , R.W. Harrison , M. Preuss , T. Ungár , P. Frankel

Protons are increasingly used as a surrogate for neutrons to study radiation damage of engineering alloys used in the core of a nuclear reactor, enabling high fluences in comparatively short times. However, the accelerated damage rate of protons is usually compensated by an increased irradiation temperature to assist diffusion. To better understand dose rate effects on microstructure evolution during radiation damage, recrystallized Low-Sn ZIRLO and Zircaloy-2 were proton-irradiated to 0.15 dpa at 320 °C using nominal dose rates of 1.3, 2.5, and 5.2 × 10⁻⁵ dpa/s. Depth profiling using microbeam synchrotron XRD was conducted across the 30 µm deep irradiated regions for line profile analysis, enabling dislocation line density determination. We found no significant difference in dislocation density among the different dose rates for Zircaloy-2 while Low-Sn ZIRLO displayed dose rate sensitive microstructural evolution. However, Low-Sn ZIRLO exhibited a significantly lower overall dislocation density compared to Zircaloy-2 samples at all dose rates. (S)TEM analysis of the samples showed clear 〈a〉 loop alignment in Zircaloy-2, while this was less pronounced in Low-Sn ZIRLO. APT analysis conducted on Low-Sn ZIRLO specimens showed the early onset of irradiation induced nanoclusters of Nb, where the clusters were observed to be comparatively smaller in the sample exposed to high dose rate irradiation. Overall, the integration of different techniques has provided a more comprehensive understanding of the early-stage damage evolution under differing damage accumulation rates.

{"title":"The effect of proton irradiation dose rate on the evolution of microstructure in Zr alloys: A synchrotron microbeam X-ray, TEM, and APT study","authors":"Ö. Koç , R. Thomas , B. Jenkins , C. Hofer , Z. Hegedüs , U. Lienert , R.W. Harrison , M. Preuss , T. Ungár , P. Frankel","doi":"10.1016/j.jnucmat.2025.155721","DOIUrl":"10.1016/j.jnucmat.2025.155721","url":null,"abstract":"<div><div>Protons are increasingly used as a surrogate for neutrons to study radiation damage of engineering alloys used in the core of a nuclear reactor, enabling high fluences in comparatively short times. However, the accelerated damage rate of protons is usually compensated by an increased irradiation temperature to assist diffusion. To better understand dose rate effects on microstructure evolution during radiation damage, recrystallized Low-Sn ZIRLO and Zircaloy-2 were proton-irradiated to 0.15 dpa at 320 °C using nominal dose rates of 1.3, 2.5, and 5.2 × 10<sup>−5</sup> dpa/s. Depth profiling using microbeam synchrotron XRD was conducted across the 30 µm deep irradiated regions for line profile analysis, enabling dislocation line density determination. We found no significant difference in dislocation density among the different dose rates for Zircaloy-2 while Low-Sn ZIRLO displayed dose rate sensitive microstructural evolution. However, Low-Sn ZIRLO exhibited a significantly lower overall dislocation density compared to Zircaloy-2 samples at all dose rates. (S)TEM analysis of the samples showed clear 〈a〉 loop alignment in Zircaloy-2, while this was less pronounced in Low-Sn ZIRLO. APT analysis conducted on Low-Sn ZIRLO specimens showed the early onset of irradiation induced nanoclusters of Nb, where the clusters were observed to be comparatively smaller in the sample exposed to high dose rate irradiation. Overall, the integration of different techniques has provided a more comprehensive understanding of the early-stage damage evolution under differing damage accumulation rates.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155721"},"PeriodicalIF":2.8,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548942","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}

引用次数: 0

Development of a dynamic-mesh porosity transport model for multi-dimensional fuel performance codes

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

Journal of Nuclear Materials

Pub Date : 2025-02-28 DOI: 10.1016/j.jnucmat.2025.155717

Edoardo Luciano Brunetto , Carlo Fiorina , Andreas Pautz , Sander van Til , Fitriana Nindiyasari , Alexander Fedorov , Alessandro Scolaro

The porosity redistribution within nuclear fuel pellets exposed to high power ratings plays a critical role in the thermo-mechanical behavior of fast reactor fuel. Traditional fuel performance codes predict porosity migration through advection-dominated transport equations often assuming a fixed geometry, and limiting their accuracy in asymmetric conditions. A novel dynamic-mesh porosity migration model has been developed to address these limitations. For verification and demonstration purposes, the model has been implemented in OFFBEAT, a multidimensional OpenFOAM-based fuel performance code. The solver dynamically adjusts the fuel pellet geometry to model the evolution of the central hole caused by pore migration. Mesh quality is preserved throughout the simulation by means dynamic-mesh algorithms involving the resolution of a mesh-motion equation to diffuse the displacement imposed at the mesh boundaries to all the domain points. The methodology incorporates modifications to the traditional porosity transport equation, correcting the advective fluxes in the governing equations to account for mesh points movement. A simple mechanistic model to determine the hole expansion velocity as a function of the local porosity, pore velocity and inner fuel radius is proposed. The model's parameters are calibrated using open literature experimental data, demonstrating the solver capability to predict central void diameters within acceptable discrepancy. The dynamic-mesh solver shows good accuracy in predicting off-centered hole formations and aligns well with post-irradiation examination data. This new approach preserves the foundational principles of existing porosity migration models while offering enhanced flexibility and accuracy in asymmetric heat transfer scenarios.

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引用次数: 0

Microstructure, oxidation kinetics and hydrogen absorption of Cr-coated Zr-Sn-Nb alloy cladding tubes after single-sided oxidation at 1000–1200 °C followed by fast reflood

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

Journal of Nuclear Materials

Pub Date : 2025-02-27 DOI: 10.1016/j.jnucmat.2025.155718

Weiwei Xiao , Sheng Xu , Xiao Hu , Jinghao Huang , Shihong Liu , Shuliang Zou

Reflood of nuclear fuel assemblies is the top priority accident management strategy for nuclear power plants in the event of a loss of coolant accident, during which the cladding tubes inevitably undergo reflood oxidation. This study aims to investigate the single-sided reflood oxidation behavior of Cr-coated Zr-Sn-Nb alloy cladding tubes at 1000 °C-1200 °C. High-temperature steam oxidation and in-situ quenching were employed to simulate the reflood oxidation process of nuclear fuel assembly cladding tubes in the early stages of severe accidents. The microstructure, cross-sectional layer thickness evolution, oxidation kinetics, and hydrogen absorption of Cr-coated Zr-Sn-Nb alloy cladding tubes during single-sided reflood oxidation process were investigated. The results showed that after single-sided reflood oxidation, microcracks appeared on the surface of the cladding tubes. As the oxidation temperature increases and the oxidation time prolongs, the surface oxidation products gradually evolve from porous flocculent structures to strip-shaped or elliptical bubble structures and worm aggregated structures. A multi-layer layered structure of Cr₂O₃ layer/Cr coating/Cr-Zr diffusion layer/α-Zr(O) was formed on the cross-section of the cladding tube after single-sided reflood oxidation. The thickness of the Cr₂O₃ layer and residual Cr coating does not increase or decrease monotonically with the extension of oxidation time after reflood oxidation at 1200 °C. The kinetics of single-sided reflood oxidation follows a parabolic law, and the oxidation constant increases by about an order of magnitude as the oxidation temperature increases by 100 °C. As the oxidation temperature increases and oxidation time prolongs, the hydrogen absorption of the cladding tube gradually increases. After single-sided reflood oxidation, the hydrides in the Zr-Sn-Nb alloy cladding tube are mainly δ-ZrH_1.5.

{"title":"Microstructure, oxidation kinetics and hydrogen absorption of Cr-coated Zr-Sn-Nb alloy cladding tubes after single-sided oxidation at 1000–1200 °C followed by fast reflood","authors":"Weiwei Xiao , Sheng Xu , Xiao Hu , Jinghao Huang , Shihong Liu , Shuliang Zou","doi":"10.1016/j.jnucmat.2025.155718","DOIUrl":"10.1016/j.jnucmat.2025.155718","url":null,"abstract":"<div><div>Reflood of nuclear fuel assemblies is the top priority accident management strategy for nuclear power plants in the event of a loss of coolant accident, during which the cladding tubes inevitably undergo reflood oxidation. This study aims to investigate the single-sided reflood oxidation behavior of Cr-coated Zr-Sn-Nb alloy cladding tubes at 1000 °C-1200 °C. High-temperature steam oxidation and in-situ quenching were employed to simulate the reflood oxidation process of nuclear fuel assembly cladding tubes in the early stages of severe accidents. The microstructure, cross-sectional layer thickness evolution, oxidation kinetics, and hydrogen absorption of Cr-coated Zr-Sn-Nb alloy cladding tubes during single-sided reflood oxidation process were investigated. The results showed that after single-sided reflood oxidation, microcracks appeared on the surface of the cladding tubes. As the oxidation temperature increases and the oxidation time prolongs, the surface oxidation products gradually evolve from porous flocculent structures to strip-shaped or elliptical bubble structures and worm aggregated structures. A multi-layer layered structure of Cr<sub>2</sub>O<sub>3</sub> layer/Cr coating/Cr-Zr diffusion layer/α-Zr(O) was formed on the cross-section of the cladding tube after single-sided reflood oxidation. The thickness of the Cr<sub>2</sub>O<sub>3</sub> layer and residual Cr coating does not increase or decrease monotonically with the extension of oxidation time after reflood oxidation at 1200 °C. The kinetics of single-sided reflood oxidation follows a parabolic law, and the oxidation constant increases by about an order of magnitude as the oxidation temperature increases by 100 °C. As the oxidation temperature increases and oxidation time prolongs, the hydrogen absorption of the cladding tube gradually increases. After single-sided reflood oxidation, the hydrides in the Zr-Sn-Nb alloy cladding tube are mainly δ-ZrH<sub>1.5</sub>.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155718"},"PeriodicalIF":2.8,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143535341","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

Cesium and iodine speciation in irradiated UO2 fuel

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

Journal of Nuclear Materials

Pub Date : 2025-02-26 DOI: 10.1016/j.jnucmat.2025.155715

J.-Y. Colle , J.N. Zappey , O. Beneš , M. Cologna , T. Wiss , R.J.M. Konings

The presence of CsI in nuclear fuel has long been debated. Its formation significantly decreases volatility, thereby reducing the rate at which iodine and cesium are released from the reactor core during a nuclear accident. A series of samples were investigated by Knudsen Effusion Mass Spectrometry (KEMS) in order to determine whether CsI is present in irradiated nuclear fuel. The examined samples were pure CsI, CsI exposed to gamma radiation, CsI-doped UO₂ simulated fuel and irradiated LWR fuel samples. The CsI and CsI-doped samples were examined to establish boundary conditions for the detection of CsI by KEMS. These samples indicated that the presence of CsI in fuel is characterized by three mass spectrometric signals Cs⁺, I⁺ and CsI⁺, with a peak ratio of CsI⁺ and I⁺ of 1:0.7. The examinations of irradiated fuels showed none of these characteristics and hence no evidence that CsI is present in irradiated LWR nuclear fuel, at least after a storage period of years.

引用次数: 0