This study introduces an improved non-dominated sorting genetic algorithm II (INSGA-II) to optimize the boundary delineation of sandstone-type uranium deposits by determining the threshold of uranium content per square meter (UCPSM, kg/m2) for minable units. A multi-objective optimization model was developed to maximize both economic and resource benefits, which was solved using the INSGA-II. Key enhancements include: (1) population initialization via symmetric Latin hypercube design (SLHD); (2) adaptive mutation and crossover parameters. Applied to real data from a Chinese mining area, the model and algorithm demonstrated practical effectiveness. The Pareto solution set derived from the optimization enabled the determination of UCPSM thresholds, supporting a novel mining-area boundary definition method based on aggregating minable units and unlocking the resource and potential economic value of idle uranium deposits. This approach offers a new decision-making tool for sandstone-type uranium mining area design.
{"title":"Techno-Economic optimization of sandstone uranium Mining: A Case study of uranium content per square meter","authors":"Jiabing Li , Chuanfei Zhang , Xiangxue Zhang , Meifang Chen , Mingtao Jia","doi":"10.1016/j.anucene.2026.112125","DOIUrl":"10.1016/j.anucene.2026.112125","url":null,"abstract":"<div><div>This study introduces an improved non-dominated sorting genetic algorithm II (INSGA-II) to optimize the boundary delineation of sandstone-type uranium deposits by determining the threshold of uranium content per square meter (UCPSM, kg/m<sup>2</sup>) for minable units. A multi-objective optimization model was developed to maximize both economic and resource benefits, which was solved using the INSGA-II. Key enhancements include: (1) population initialization via symmetric Latin hypercube design (SLHD); (2) adaptive mutation and crossover parameters. Applied to real data from a Chinese mining area, the model and algorithm demonstrated practical effectiveness. The Pareto solution set derived from the optimization enabled the determination of UCPSM thresholds, supporting a novel mining-area boundary definition method based on aggregating minable units and unlocking the resource and potential economic value of idle uranium deposits. This approach offers a new decision-making tool for sandstone-type uranium mining area design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112125"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035579","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2026-01-20DOI: 10.1016/j.anucene.2026.112158
Xinling Dai , Dechang Cai , Jin Cai
Criticality safety analysis is essential for fuel assembly transport, as it ensures subcriticality under all potential accident scenarios. Traditional methods are computationally expensive, requiring hundreds of input cases. The Neutron Worth Iteration method was developed to efficiently determine conservative envelopes for PWR fuel assemblies under transport impacts. By iteratively adjusting the configuration of fuel rods according to the neutron worth distribution within a fuel assembly, the method optimizes rod positioning to maximize . This method significantly reduces computational effort while maintaining a slight deviation of less than 2.62. This approach enables rapid yet rigorous criticality safety assessments.
{"title":"Efficient critical safety analysis model for PWR fuel assembly under transport accidents by neutron worth iteration","authors":"Xinling Dai , Dechang Cai , Jin Cai","doi":"10.1016/j.anucene.2026.112158","DOIUrl":"10.1016/j.anucene.2026.112158","url":null,"abstract":"<div><div>Criticality safety analysis is essential for fuel assembly transport, as it ensures subcriticality under all potential accident scenarios. Traditional methods are computationally expensive, requiring hundreds of input cases. The Neutron Worth Iteration method was developed to efficiently determine conservative <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span> envelopes for PWR fuel assemblies under transport impacts. By iteratively adjusting the configuration of fuel rods according to the neutron worth distribution within a fuel assembly, the method optimizes rod positioning to maximize <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span>. This method significantly reduces computational effort while maintaining a slight <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span> deviation of less than 2.62<span><math><mo>‰</mo></math></span>. This approach enables rapid yet rigorous criticality safety assessments.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112158"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035575","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2026-01-20DOI: 10.1016/j.anucene.2026.112141
H.C. Manjunatha , B.M. Sankarshan , P.S. Damodara Gupta , L. Seenappa , K.N. Sridhar , R. Munirathnam
This study investigates alternative materials to lead for radiation shielding, addressing the need for safer and more effective options. Traditional materials like lead, although effective due to their high atomic number, are toxic and pose environmental risks. The study explores a set of tantalum–tungsten–oxygen (Ta–W–O) compounds, including TaWO, TaWO, TaWO, TaWO, and others. These compounds offer promising shielding properties due to their high density, atomic number, and stability. Key shielding parameters such as mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half-value layer (HVL), and effective atomic number (Z) were calculated and compared to lead. Among all the studied Ta–W–O compounds, TaWO3 was identified as the most efficient and thermodynamically stable lead-free shielding material, exhibiting the highest photon attenuation performance across low- and intermediate-energy ranges. Across various energy ranges, these compounds demonstrate superior radiation protection efficiency (RPE) and electron density, essential for shielding in healthcare, nuclear, and aerospace applications. The findings suggest that tantalum–tungsten compounds could serve as viable lead-free shielding materials, offering a safer and more sustainable alternative for radiation protection.
{"title":"Evaluation of tantalum–tungsten–oxygen compounds as lead-free radiation shielding materials","authors":"H.C. Manjunatha , B.M. Sankarshan , P.S. Damodara Gupta , L. Seenappa , K.N. Sridhar , R. Munirathnam","doi":"10.1016/j.anucene.2026.112141","DOIUrl":"10.1016/j.anucene.2026.112141","url":null,"abstract":"<div><div>This study investigates alternative materials to lead for radiation shielding, addressing the need for safer and more effective options. Traditional materials like lead, although effective due to their high atomic number, are toxic and pose environmental risks. The study explores a set of tantalum–tungsten–oxygen (Ta–W–O) compounds, including TaW<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<span><math><msub><mrow></mrow><mrow><mn>3</mn></mrow></msub></math></span>, TaW<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>, Ta<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>W<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<span><math><msub><mrow></mrow><mrow><mn>5</mn></mrow></msub></math></span>, TaWO<span><math><msub><mrow></mrow><mrow><mn>3</mn></mrow></msub></math></span>, and others. These compounds offer promising shielding properties due to their high density, atomic number, and stability. Key shielding parameters such as mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half-value layer (HVL), and effective atomic number (Z<span><math><msub><mrow></mrow><mrow><mtext>eff</mtext></mrow></msub></math></span>) were calculated and compared to lead. Among all the studied Ta–W–O compounds, TaW<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<sub>3</sub> was identified as the most efficient and thermodynamically stable lead-free shielding material, exhibiting the highest photon attenuation performance across low- and intermediate-energy ranges. Across various energy ranges, these compounds demonstrate superior radiation protection efficiency (RPE) and electron density, essential for shielding in healthcare, nuclear, and aerospace applications. The findings suggest that tantalum–tungsten compounds could serve as viable lead-free shielding materials, offering a safer and more sustainable alternative for radiation protection.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112141"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035577","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2026-01-20DOI: 10.1016/j.anucene.2026.112153
Yilong Li , Yan Guo , Junyong Liu, Hongyu Yang, Guangyun Min, Naibin Jiang
This study aims to investigate the flow disturbance effect of different spacer grids in a Pressurized Water Reactor (PWR) and their influence on the flow-induced vibration characteristics of fuel rods. Based on the model from the Subchannel and Bundle Test (PSBT). Differences between two-way flow-structure interaction and one-way flow-structure interaction analyzed. Subsequently, the one-way flow-structure interaction method was adopted. Under the condition of consistent average inlet mean flow velocity, compared the results of each spacer grid subchannel: transverse flow, pressure distribution, and pressure drop. These effects show significant correlation with grid structural features such as rigid protrusion, spring, and mixing vanes. Thus amplitude are significant in subchannels: the influence of the simple supported grid is negligible, spacer grids with no mixing vanes has a significant influence, and spacer grids with mixing vanes have the most significant effects.
{"title":"Numerical investigation of spacer grid-induced flow disturbances and impact on fuel rod flow-induced vibrations","authors":"Yilong Li , Yan Guo , Junyong Liu, Hongyu Yang, Guangyun Min, Naibin Jiang","doi":"10.1016/j.anucene.2026.112153","DOIUrl":"10.1016/j.anucene.2026.112153","url":null,"abstract":"<div><div>This study aims to investigate the flow disturbance effect of different spacer grids in a Pressurized Water Reactor (PWR) and their influence on the flow-induced vibration characteristics of fuel rods. Based on the model from the Subchannel and Bundle Test (PSBT). Differences between two-way flow-structure interaction and one-way flow-structure interaction analyzed. Subsequently, the one-way flow-structure interaction method was adopted. Under the condition of consistent average inlet mean flow velocity, compared the results of each spacer grid subchannel: transverse flow, pressure distribution, and pressure drop. These effects show significant correlation with grid structural features such as rigid protrusion, spring, and mixing vanes. Thus amplitude are significant in subchannels: the influence of the simple supported grid is negligible, spacer grids with no mixing vanes has a significant influence, and spacer grids with mixing vanes have the most significant effects.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112153"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035578","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2026-01-20DOI: 10.1016/j.anucene.2026.112147
Longji Qiu, Nan Chao, Yong-kuo Liu, Zongzhen Shi, Joseph Daniel
This paper proposes a gamma radiation field reconstruction method based on sparse detection data. By integrating compressed sensing (CS) theory with clustered transport theory, a sensing matrix for radiation field reconstruction is constructed. A novel Non-negative Lasso Orthogonal Matching Pursuit (NNLasso-OMP) algorithm is developed, combining the high efficiency and flexibility of OMP with the overfitting resistance of Lasso. To evaluate the proposed method, three simulation scenarios are conducted, using Monte Carlo simulation results as reference benchmarks. The reconstruction performance of NNLasso-OMP is compared with that of OMP, inverse distance weighting (IDW) and 3DCNN algorithms. Results show that the average relative error (ARE) of NNLasso-OMP remains below 10% across all scenarios, achieving a reconstruction success rate (SR) exceeding 95%, while accurately identifying source locations. The proposed NNLasso-OMP method outperforms both OMP and IDW, demonstrating its effectiveness in achieving high-quality gamma radiation field reconstruction from sparse measurements.
{"title":"A non-negative Lasso Orthogonal Matching Pursuit method for gamma radiation field reconstruction with sparse measurement data","authors":"Longji Qiu, Nan Chao, Yong-kuo Liu, Zongzhen Shi, Joseph Daniel","doi":"10.1016/j.anucene.2026.112147","DOIUrl":"10.1016/j.anucene.2026.112147","url":null,"abstract":"<div><div>This paper proposes a gamma radiation field reconstruction method based on sparse detection data. By integrating compressed sensing (CS) theory with clustered transport theory, a sensing matrix for radiation field reconstruction is constructed. A novel Non-negative Lasso Orthogonal Matching Pursuit (NNLasso-OMP) algorithm is developed, combining the high efficiency and flexibility of OMP with the overfitting resistance of Lasso. To evaluate the proposed method, three simulation scenarios are conducted, using Monte Carlo simulation results as reference benchmarks. The reconstruction performance of NNLasso-OMP is compared with that of OMP, inverse distance weighting (IDW) and 3DCNN algorithms. Results show that the average relative error (ARE) of NNLasso-OMP remains below 10% across all scenarios, achieving a reconstruction success rate (SR) exceeding 95%, while accurately identifying source locations. The proposed NNLasso-OMP method outperforms both OMP and IDW, demonstrating its effectiveness in achieving high-quality gamma radiation field reconstruction from sparse measurements.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112147"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035571","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2026-01-20DOI: 10.1016/j.anucene.2025.112009
Restu Kojo
This study evaluates the thermal-hydraulics in the containment of boiling water reactors caused by superheated gas leakage under severe accident conditions. The research involved characterizing the heat transfer paths, which included identifying superheated gas leakage positions and heat release from the reactor pressure vessel boundaries, classifying the containment volumes, and categorizing potential failure sections of the containment. Based on these insights, the heat transfer paths during severe accidents were clarified, and accident scenarios considering leakage from the safety relief valve and traversing in-core probe tubes were selected as representative scenarios.
A three-dimensional computational fluid dynamics (CFD) model of the Mark I containment was developed to evaluate the thermal-hydraulics of an entire drywell of the containment. Special attention was given to modeling the detailed structures of the containment top head flange, radiation from the reactor pressure vessel upper head, condensation in the suppression pool, and heat release from the containment top head to the reactor well. The CFD analyses focused on two scenarios: safety relief valve leakage and traversing in-core probe tube leakage, which can result in significant temperature distribution in the upper and lower drywell, respectively. This study identified the high temperature location on the containment boundary with higher possibility of failure by high-temperature gas leakage under severe accident conditions based on temperature distribution obtained by the present detailed three-dimensional CFD analysis.
{"title":"Identification of the containment heating mechanism and temperature distribution by high-temperature gas leakage under severe accident conditions","authors":"Restu Kojo","doi":"10.1016/j.anucene.2025.112009","DOIUrl":"10.1016/j.anucene.2025.112009","url":null,"abstract":"<div><div>This study evaluates the thermal-hydraulics in the containment of boiling water reactors caused by superheated gas leakage under severe accident conditions. The research involved characterizing the heat transfer paths, which included identifying superheated gas leakage positions and heat release from the reactor pressure vessel boundaries, classifying the containment volumes, and categorizing potential failure sections of the containment. Based on these insights, the heat transfer paths during severe accidents were clarified, and accident scenarios considering leakage from the safety relief valve and traversing in-core probe tubes were selected as representative scenarios.</div><div>A three-dimensional computational fluid dynamics (CFD) model of the Mark I containment was developed to evaluate the thermal-hydraulics of an entire drywell of the containment. Special attention was given to modeling the detailed structures of the containment top head flange, radiation from the reactor pressure vessel upper head, condensation in the suppression pool, and heat release from the containment top head to the reactor well. The CFD analyses focused on two scenarios: safety relief valve leakage and traversing in-core probe tube leakage, which can result in significant temperature distribution in the upper and lower drywell, respectively. This study identified the high temperature location on the containment boundary with higher possibility of failure by high-temperature gas leakage under severe accident conditions based on temperature distribution obtained by the present detailed three-dimensional CFD analysis.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112009"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035576","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2026-01-19DOI: 10.1016/j.anucene.2026.112119
Yunxiang Li , Runsheng Yang , Yuefeng Guo , Xingkang Su , Yuping Zhou , Jian Hong , Yuxing Liu , Zinan Huang , Xin Su , Youpeng Zhang , WenJun Hu , Long Gu
The study of the thermo-hydraulic behavior of liquid lead–bismuth in wire-wrapped fuel assemblies is of great significance for the safety design of the CiADS subcritical reactor. The four-equation turbulence model, which introduces both dynamic and thermal turbulence time scales to transport the turbulent Prandtl number, offers improved numerical accuracy in simulating the heat transfer of liquid lead–bismuth eutectic. To further enhance the heat-transfer performance of LBE within the fuel assembly, the arrangement and rotational direction of the spacer wires in the lead-cooled fast reactor fuel assembly were optimized, and a directionally-alternated wire-wrapped fuel assembly was proposed. In this work, a customized CFD solver named LBE4EqnFoam was developed based on the open-source platform OpenFOAM. Using the developed solver, detailed simulations were conducted for both the conventional wire-wrapped fuel assembly and the directionally-alternated wire-wrapped fuel assembly. The results show that LBE4EqnFoam provides highly accurate predictions of LBE flow and heat transfer in complex geometries. The solver has been validated against experimental measurements, showing that the maximum relative error in predicting the coolant temperature is below 2%, while the maximum relative error in predicting the cladding surface temperature is below 3%. Compared with the conventional design, the maximum pressure-drop reduction achieved by the directionally-alternated wire-wrapped fuel assembly is 28.22%, and the reduction at the outlet is 25.56%, which helps decrease the required pump head and the measurement range of pressure sensors. The directionally-alternated configuration also enhances cross-mixing among subchannels, leading to a more uniform temperature field and smaller temperature gradients at the outlet. This improvement is beneficial for reducing thermal fatigue and creep risks in the structural components near the outlet region. Moreover, the directionally-alternated design achieves a global average Nusselt number that is 1.38 times that of the conventional configuration. Furthermore, the directionally-alternated wire-wrapped fuel assembly exhibits superior integrated heat-transfer performance, with the integrated thermal–hydraulic factor improved by 36.45% compared with the conventional configuration.
{"title":"Numerical study on the thermo-hydraulic behaviors of the Directionally-Alternated wire-wrapped fuel assembly in lead-cooled fast reactors based on SSTSAS k-ω-kθ-εθ four-equation model","authors":"Yunxiang Li , Runsheng Yang , Yuefeng Guo , Xingkang Su , Yuping Zhou , Jian Hong , Yuxing Liu , Zinan Huang , Xin Su , Youpeng Zhang , WenJun Hu , Long Gu","doi":"10.1016/j.anucene.2026.112119","DOIUrl":"10.1016/j.anucene.2026.112119","url":null,"abstract":"<div><div>The study of the thermo-hydraulic behavior of liquid lead–bismuth in wire-wrapped fuel assemblies is of great significance for the safety design of the CiADS subcritical reactor. The four-equation turbulence model, which introduces both dynamic and thermal turbulence time scales to transport the turbulent Prandtl number, offers improved numerical accuracy in simulating the heat transfer of liquid lead–bismuth eutectic. To further enhance the heat-transfer performance of LBE within the fuel assembly, the arrangement and rotational direction of the spacer wires in the lead-cooled fast reactor fuel assembly were optimized, and a directionally-alternated wire-wrapped fuel assembly was proposed. In this work, a customized CFD solver named LBE4EqnFoam was developed based on the open-source platform OpenFOAM. Using the developed solver, detailed simulations were conducted for both the conventional wire-wrapped fuel assembly and the directionally-alternated wire-wrapped fuel assembly. The results show that LBE4EqnFoam provides highly accurate predictions of LBE flow and heat transfer in complex geometries. The solver has been validated against experimental measurements, showing that the maximum relative error in predicting the coolant temperature is below 2%, while the maximum relative error in predicting the cladding surface temperature is below 3%. Compared with the conventional design, the maximum pressure-drop reduction achieved by the directionally-alternated wire-wrapped fuel assembly is 28.22%, and the reduction at the outlet is 25.56%, which helps decrease the required pump head and the measurement range of pressure sensors. The directionally-alternated configuration also enhances cross-mixing among subchannels, leading to a more uniform temperature field and smaller temperature gradients at the outlet. This improvement is beneficial for reducing thermal fatigue and creep risks in the structural components near the outlet region. Moreover, the directionally-alternated design achieves a global average Nusselt number that is 1.38 times that of the conventional configuration. Furthermore, the directionally-alternated wire-wrapped fuel assembly exhibits superior integrated heat-transfer performance, with the integrated thermal–hydraulic factor improved by 36.45% compared with the conventional configuration.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112119"},"PeriodicalIF":2.3,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035572","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2026-01-19DOI: 10.1016/j.anucene.2026.112148
Kaiyuan Wang, Yayun Luo, Yanan Zhu, Xin Jin
Cr2O3-doped large-grain UO2 pellets enhance fission gas retention and fuel rod accident tolerance. In this paper, drawing upon the behavior mechanism of fission gas, a mechanistic fission gas release model has been developed to evaluate fission gas release in the engineering design of large-grained UO2 fuel rods. The model integrates diffusion coefficients derived from free-energy cluster dynamics and couples with JASMINE from the NATENE package. Based on the new model-based software, a detailed validation was performed by combining the experimental data from international projects. The results show excellent agreement between simulation results and experimental measurements for key parameters, including fission gas release rate, internal pressure, and temperature. A prediction deviation of the fission gas release rate remains within ±35%. In summary, the new model-based software demonstrates the capability to simulate in-reactor fission gas behavior for Cr2O3-doped large-grained UO2 nuclear fuel.
{"title":"Development and verification of a new fission gas release model for large-grained UO2 pellets","authors":"Kaiyuan Wang, Yayun Luo, Yanan Zhu, Xin Jin","doi":"10.1016/j.anucene.2026.112148","DOIUrl":"10.1016/j.anucene.2026.112148","url":null,"abstract":"<div><div>Cr<sub>2</sub>O<sub>3</sub>-doped large-grain UO<sub>2</sub> pellets enhance fission gas retention and fuel rod accident tolerance. In this paper, drawing upon the behavior mechanism of fission gas, a mechanistic fission gas release model has been developed to evaluate fission gas release in the engineering design of large-grained UO<sub>2</sub> fuel rods. The model integrates diffusion coefficients derived from free-energy cluster dynamics and couples with JASMINE from the NATENE package. Based on the new model-based software, a detailed validation was performed by combining the experimental data from international projects. The results show excellent agreement between simulation results and experimental measurements for key parameters, including fission gas release rate, internal pressure, and temperature. A prediction deviation of the fission gas release rate remains within ±35%. In summary, the new model-based software demonstrates the capability to simulate in-reactor fission gas behavior for Cr<sub>2</sub>O<sub>3</sub>-doped large-grained UO<sub>2</sub> nuclear fuel.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112148"},"PeriodicalIF":2.3,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035570","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The application of liquid lead–bismuth eutectic (LBE) alloy coolant technology necessitates the implementation of real-time monitoring of dissolved oxygen concentration. Furthermore, in accordance with operational requirements, the dissolved oxygen concentration within the liquid LBE must be maintained within a reasonable target range. In order to enhance the provision of rapid and efficient oxygen replenishment to the liquid LBE loop, the oxygen supply behavior of the mass exchanger (MX) was modelled. The development of the oxygen control model of the MX was achieved by the collection of input (MX temperature) and output (signal of oxygen sensor) data from the solid-phase oxygen control experiments in the liquid LBE recirculation loop. The least squares method and neural network algorithm were utilised in the development of the oxygen control model, respectively. The findings demonstrate the efficacy of the oxygen control model in predicting the dissolved oxygen concentration within the LBE loop. This provides a theoretical framework for the subsequent optimisation of the solid-phase oxygen control strategy within the LBE system.
{"title":"Dissolved oxygen concentration control and prediction modelling for liquid LBE loop: UPBEAT loop","authors":"Ruixian Liang, Wei Mao, Xiangtian Hou, Zulong Hao, Haicai Lyu, Hao Wu, Huiping Zhu, Fang Liu, Yang Liu, Fenglei Niu","doi":"10.1016/j.anucene.2026.112140","DOIUrl":"10.1016/j.anucene.2026.112140","url":null,"abstract":"<div><div>The application of liquid lead–bismuth eutectic (LBE) alloy coolant technology necessitates the implementation of real-time monitoring of dissolved oxygen concentration. Furthermore, in accordance with operational requirements, the dissolved oxygen concentration within the liquid LBE must be maintained within a reasonable target range. In order to enhance the provision of rapid and efficient oxygen replenishment to the liquid LBE loop, the oxygen supply behavior of the mass exchanger (MX) was modelled. The development of the oxygen control model of the MX was achieved by the collection of input (MX temperature) and output (signal of oxygen sensor) data from the solid-phase oxygen control experiments in the liquid LBE recirculation loop. The least squares method and neural network algorithm were utilised in the development of the oxygen control model, respectively. The findings demonstrate the efficacy of the oxygen control model in predicting the dissolved oxygen concentration within the LBE loop. This provides a theoretical framework for the subsequent optimisation of the solid-phase oxygen control strategy within the LBE system.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112140"},"PeriodicalIF":2.3,"publicationDate":"2026-01-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035568","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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 : 2026-01-17DOI: 10.1016/j.anucene.2026.112127
Xinwu Su , Yongli Xu , Yinlu Han
For research on fission or fusion nuclear reactor systems, there is a pressing need for neutron reaction data on Am at incident energies up to 200 MeV. A consistent evaluation and calculation of nuclear data for n + Am reactions below 200 MeV have been performed using theoretical models, including the optical model, distorted-wave Born approximation (DWBA), Hauser–Feshbach theory with width fluctuation correction, fission model, evaporation model, exciton model, and the intranuclear cascade model. Furthermore, newly available experimental data have been incorporated. The theoretical predictions are compared with experimental measurements, as well as with evaluated data from ENDF/B-VIII.1 and JENDL-5.
{"title":"Neutron Data Evaluation in the n + 241,243Am reactions below 200 MeV","authors":"Xinwu Su , Yongli Xu , Yinlu Han","doi":"10.1016/j.anucene.2026.112127","DOIUrl":"10.1016/j.anucene.2026.112127","url":null,"abstract":"<div><div>For research on fission or fusion nuclear reactor systems, there is a pressing need for neutron reaction data on <span><math><msup><mrow></mrow><mrow><mn>241</mn><mo>,</mo><mn>243</mn></mrow></msup></math></span>Am at incident energies up to 200 MeV. A consistent evaluation and calculation of nuclear data for n + <span><math><msup><mrow></mrow><mrow><mn>241</mn><mo>,</mo><mn>243</mn></mrow></msup></math></span>Am reactions below 200 MeV have been performed using theoretical models, including the optical model, distorted-wave Born approximation (DWBA), Hauser–Feshbach theory with width fluctuation correction, fission model, evaporation model, exciton model, and the intranuclear cascade model. Furthermore, newly available experimental data have been incorporated. The theoretical predictions are compared with experimental measurements, as well as with evaluated data from ENDF/B-VIII.1 and JENDL-5.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112127"},"PeriodicalIF":2.3,"publicationDate":"2026-01-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145975289","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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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