Pub Date : 2025-04-24DOI: 10.1016/j.anucene.2025.111463
Andrew Zillmer, Clinton Wilson, Jill Mitchell, Austen Fradeneck, Ryan Marlow, Erik Rosvall, William Green, Justin Lower
The mission of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) focuses on creating irradiation facilities for nuclear materials and fuels research. While radioisotope production is not a core part of the ATR mission, it is important for providing a U.S. domestic source of critical radioisotopes such as cobalt-60 (Co-60). This paper gives an overview for radioisotopes currently being produced in ATR as well as other potential radioisotopes that can be produced in ATR. It also provides guidance on radioisotope production that can be applied to other test reactors.
{"title":"Radioisotope production at the advanced test reactor: process and lessons learned","authors":"Andrew Zillmer, Clinton Wilson, Jill Mitchell, Austen Fradeneck, Ryan Marlow, Erik Rosvall, William Green, Justin Lower","doi":"10.1016/j.anucene.2025.111463","DOIUrl":"10.1016/j.anucene.2025.111463","url":null,"abstract":"<div><div>The mission of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) focuses on creating irradiation facilities for nuclear materials and fuels research. While radioisotope production is not a core part of the ATR mission, it is important for providing a U.S. domestic source of critical radioisotopes such as cobalt-60 (Co-60). This paper gives an overview for radioisotopes currently being produced in ATR as well as other potential radioisotopes that can be produced in ATR. It also provides guidance on radioisotope production that can be applied to other test reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111463"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863595","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 : 2025-04-24DOI: 10.1016/j.anucene.2025.111492
Yipeng Fan , Xiya Zhu , Jun Yang , Hongyun Xie
Emergency preparedness and response (EPR) are essential to protect the public health and environmental safety in the event of a nuclear or radiological emergency. In the paper, the fundamental concepts, framework and methods for nuclear and radiological emergency management are reviewed from a comprehensive perspective. Prior to the emergency classifications, the organizational framework, responsibilities and initiative programs especially undertaken by the International Atomic Energy Agency (IAEA) is literally introduced. Then the generic principles for classification of emergency planning zones (EPZ) and emergency action levels (EALs) are in detail discussed as a basis for the planning of protective actions that the national authorities and international agencies need to perform during a nuclear or radiological emergency. Meanwhile, the methodologies and tools used for radionuclides release, transport, dispersion and deposition, dose assessment are also systematically investigated and compared to provide a guide for practitioner’s research and vision in emergency preparedness planning. Finally, the major challenges and development trends associated with emergency preparedness and response are prospected with the revolution of edge artificial intelligence (AI-driven) solutions to leverage collaborative perception and risk-informed decision making in nuclear and radiological emergency management.
{"title":"A review on nuclear emergency preparedness and response management","authors":"Yipeng Fan , Xiya Zhu , Jun Yang , Hongyun Xie","doi":"10.1016/j.anucene.2025.111492","DOIUrl":"10.1016/j.anucene.2025.111492","url":null,"abstract":"<div><div>Emergency preparedness and response (EPR) are essential to protect the public health and environmental safety in the event of a nuclear or radiological emergency. In the paper, the fundamental concepts, framework and methods for nuclear and radiological emergency management are reviewed from a comprehensive perspective. Prior to the emergency classifications, the organizational framework, responsibilities and initiative programs especially undertaken by the International Atomic Energy Agency (IAEA) is literally introduced. Then the generic principles for classification of emergency planning zones (EPZ) and emergency action levels (EALs) are in detail discussed as a basis for the planning of protective actions that the national authorities and international agencies need to perform during a nuclear or radiological emergency. Meanwhile, the methodologies and tools used for radionuclides release, transport, dispersion and deposition, dose assessment are also systematically investigated and compared to provide a guide for practitioner’s research and vision in emergency preparedness planning. Finally, the major challenges and development trends associated with emergency preparedness and response are prospected with the revolution of edge artificial intelligence (AI-driven) solutions to leverage collaborative perception and risk-informed decision making in nuclear and radiological emergency management.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111492"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863596","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 : 2025-04-23DOI: 10.1016/j.anucene.2025.111505
Qianni Duan , Wei Li , Kun Zhang , Junmei Wu , Zhifeng Li
The Helical Cruciform Fuel (HCF) is an advanced Innovative Fuel Design (IFD) that present greater challenges for three-dimensional(3D) refined burnup characteristic analysis. A burnup region division method for HCF was proposed and the burnup characteristic analysis was conducted using OpenMC code. 3D refined distribution of thermal and fast neutron flux and power density, isotope densities of typical nuclides at different positions of HCF were calculated. The influences of twist pitch, axial enrichment, and boundary conditions were studied. Meanwhile, burnup characteristic on HCF and cylindrical fuel was compared with the same material and volume. The results show that, the chain reaction of HCF can last longer, and has obvious economic advantages than cylinder fuel. Unlike cylindrical fuel, the circumferential distribution of neutron physical variables for HCF is inhomogeneous and becomes more pronounced with burnup. The proposed analysis method is applicable to any specially shaped fuel and effectively predicts the neutron physical properties during burnup, forming the foundation for IFD and safety analysis.
{"title":"Three-dimensional refined burnup characteristics analysis of Helical Cruciform Fuel","authors":"Qianni Duan , Wei Li , Kun Zhang , Junmei Wu , Zhifeng Li","doi":"10.1016/j.anucene.2025.111505","DOIUrl":"10.1016/j.anucene.2025.111505","url":null,"abstract":"<div><div>The Helical Cruciform Fuel (HCF) is an advanced Innovative Fuel Design (IFD) that present greater challenges for three-dimensional(3D) refined burnup characteristic analysis. A burnup region division method for HCF was proposed and the burnup characteristic analysis was conducted using OpenMC code. 3D refined distribution of thermal and fast neutron flux and power density, isotope densities of typical nuclides at different positions of HCF were calculated. The influences of twist pitch, axial enrichment, and boundary conditions were studied. Meanwhile, burnup characteristic on HCF and cylindrical fuel was compared with the same material and volume. The results show that, the chain reaction of HCF can last longer, and has obvious economic advantages than cylinder fuel. Unlike cylindrical fuel, the circumferential distribution of neutron physical variables for HCF is inhomogeneous and becomes more pronounced with burnup. The proposed analysis method is applicable to any specially shaped fuel and effectively predicts the neutron physical properties during burnup, forming the foundation for IFD and safety analysis.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111505"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860738","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 : 2025-04-23DOI: 10.1016/j.anucene.2025.111507
Li He , Guangyao Sun , Yican Wu
High-performance compact nuclear facilities require radiation shielding designs that balance safety and weight considerations. Current intelligent shielding design methods typically combine Evolutionary Algorithms (EA) for optimization with neural networks for evaluation. However, the neural networks used, primarily BP or DNN models composed of Fully Connected (FC) layers, require large datasets and extensive computation resources. A novel method fusing Self-Attention-based Sequence Prediction Network (Self-Attention-SPN) and Non-dominated Sorting Genetic Algorithm III (NSGA-III) was proposed in this paper for multi-objective radiation shielding design optimization. By reformulating dose rate calculation as a sequence prediction problem, the SPN of lightweight network structure leverages the multi-physics feature projection and multi-head self-attention mechanism to effectively capture the inter-layer physical feature relationships, ensuring high prediction accuracy with small datasets. The method is validated using the Savannah reactor case, where SPN achieves Monte Carlo (MC)-level accuracy with significantly reduced computational cost. Comparative experiments show that training data with additional physical parameters can reduce SPN training loss, underscoring the importance of physical information. Furthermore, SPN outperforms BP in prediction accuracy, validating the effectiveness of the multi-head self-attention mechanism. Sensitivity analysis of NSGA-III coupled with SPN prediction perturbation confirms the robustness of the proposed method. The optimization solutions effectively converge to the Pareto front, demonstrating the method’s efficiency and reliability for multi-objective radiation shielding design.
{"title":"A method fusing self-attention-SPN and NSGA-III for multi-objective radiation shielding design optimization","authors":"Li He , Guangyao Sun , Yican Wu","doi":"10.1016/j.anucene.2025.111507","DOIUrl":"10.1016/j.anucene.2025.111507","url":null,"abstract":"<div><div>High-performance compact nuclear facilities require radiation shielding designs that balance safety and weight considerations. Current intelligent shielding design methods typically combine Evolutionary Algorithms (EA) for optimization with neural networks for evaluation. However, the neural networks used, primarily BP or DNN models composed of Fully Connected (FC) layers, require large datasets and extensive computation resources. A novel method fusing Self-Attention-based Sequence Prediction Network (Self-Attention-SPN) and Non-dominated Sorting Genetic Algorithm III (NSGA-III) was proposed in this paper for multi-objective radiation shielding design optimization. By reformulating dose rate calculation as a sequence prediction problem, the SPN of lightweight network structure leverages the multi-physics feature projection and multi-head self-attention mechanism to effectively capture the inter-layer physical feature relationships, ensuring high prediction accuracy with small datasets. The method is validated using the Savannah reactor case, where SPN achieves Monte Carlo (MC)-level accuracy with significantly reduced computational cost. Comparative experiments show that training data with additional physical parameters can reduce SPN training loss, underscoring the importance of physical information. Furthermore, SPN outperforms BP in prediction accuracy, validating the effectiveness of the multi-head self-attention mechanism. Sensitivity analysis of NSGA-III coupled with SPN prediction perturbation confirms the robustness of the proposed method. The optimization solutions effectively converge to the Pareto front, demonstrating the method’s efficiency and reliability for multi-objective radiation shielding design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111507"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863597","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 : 2025-04-23DOI: 10.1016/j.anucene.2025.111468
Teng Xie, Xinhua Yan, Menglu Qin, Lu Sun, Zhenfeng Tong
A cluster dynamics simulation based on rate theory was conducted to investigate the evolution of irradiation-induced dislocation loops in pure molybdenum () irradiated with + at . The changes in number density and size of dislocation loops were analyzed, revealing rapid saturation of loop density with increasing dose, while the average loop radius grows and stabilizes at a maximum size. During the initial irradiation stage, significant nucleation leads to a rapid increase in loop density. The simulation results align well with in-situ transmission electron microscopy observations and clarify the kinetic mechanisms underlying key experimental features. It was found that the diffusion coefficients and thermal equilibrium concentrations of interstitial atoms and vacancies play critical roles in determining the evolution of loop density and size at elevated temperatures. This study provides insight into the microstructural evolution of irradiated molybdenum under helium ion exposure.
{"title":"Cluster-dynamics study on the evolution of dislocation loops in molybdenum: The role of defect diffusion coefficient and equilibrium concentration","authors":"Teng Xie, Xinhua Yan, Menglu Qin, Lu Sun, Zhenfeng Tong","doi":"10.1016/j.anucene.2025.111468","DOIUrl":"10.1016/j.anucene.2025.111468","url":null,"abstract":"<div><div>A cluster dynamics simulation based on rate theory was conducted to investigate the evolution of irradiation-induced dislocation loops in pure molybdenum (<span><math><mi>Mo</mi></math></span>) irradiated with <span><math><mrow><mn>30</mn><mspace></mspace><mi>keV</mi></mrow></math></span> <span><math><mi>He</mi></math></span><sup>+</sup> at <span><math><mrow><mn>673</mn><mspace></mspace><mi>K</mi></mrow></math></span>. The changes in number density and size of dislocation loops were analyzed, revealing rapid saturation of loop density with increasing dose, while the average loop radius grows and stabilizes at a maximum size. During the initial irradiation stage, significant nucleation leads to a rapid increase in loop density. The simulation results align well with in-situ transmission electron microscopy observations and clarify the kinetic mechanisms underlying key experimental features. It was found that the diffusion coefficients and thermal equilibrium concentrations of interstitial atoms and vacancies play critical roles in determining the evolution of loop density and size at elevated temperatures. This study provides insight into the microstructural evolution of irradiated molybdenum under helium ion exposure.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111468"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863454","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 scattering anisotropy multigroup cross-section (MGXS) with respect to the outgoing angle, as well as the anisotropy of the total MGXS with respect to the incident angle, are critical for generating high-precision MGXS used in transport core calculations. A TC-MHT method based on continuous energy Monte-Carlo is developed to generate high-accuracy P0 MGXS. This method addresses scattering anisotropy using transport correction (TC) and handles the anisotropy of total cross-section through the flux-moment homogenization technique (MHT). This method is validated using the Megapower benchmark. In reactivity calculations, the bias is decomposed and the reasons for the bias are analyzed. The bias is reduced to less than 160 pcm by using the TC-MHT method. For calculations of control drum worth, the bias does not exceed 0.2%. In power calculations, the maximum pin-by-pin power distribution deviation is within 4%. This demonstrates that TC-MHT method can generate high-precision, highly adaptable MGXS for deterministic code.
{"title":"Transport-corrected flux-moment homogenization method for generating P0 multigroup cross-section based on continuous energy Monte-Carlo","authors":"Yuyang Shen , Yiwei Wu , Qufei Song , Kuaiyuan Feng , Hui Guo , Hanyang Gu","doi":"10.1016/j.anucene.2025.111448","DOIUrl":"10.1016/j.anucene.2025.111448","url":null,"abstract":"<div><div>The scattering anisotropy multigroup cross-section (MGXS) with respect to the outgoing angle, as well as the anisotropy of the total MGXS with respect to the incident angle, are critical for generating high-precision MGXS used in transport core calculations. A TC-MHT method based on continuous energy Monte-Carlo is developed to generate high-accuracy P<sub>0</sub> MGXS. This method addresses scattering anisotropy using transport correction (TC) and handles the anisotropy of total cross-section through the flux-moment homogenization technique (MHT). This method is validated using the Megapower benchmark. In reactivity calculations, the bias is decomposed and the reasons for the bias are analyzed. The bias is reduced to less than 160 pcm by using the TC-MHT method. For calculations of control drum worth, the bias does not exceed 0.2%. In power calculations, the maximum pin-by-pin power distribution deviation is within 4%. This demonstrates that TC-MHT method can generate high-precision, highly adaptable MGXS for deterministic code.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111448"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860726","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 : 2025-04-23DOI: 10.1016/j.anucene.2025.111467
Henglong Lin , Zeguang Li , Kan Wang , Weilin Zhuge , Ganglin Yu , Hongsheng Jiang , Jingkang Li , Jun Yang , Zilin Su , Huifu Wang
Heat pipe cooled reactor (HPR) has the characteristics of small size and high safety, and can be used in special application scenarios such as vehicle-mounted mobility. In this paper, the Mobile Nuclear Power System based on HPR named MNPS-1000, which is designed to generate 1MWe, is selected for research and analysis. According to the design, a whole system model is established, which includes the reactor model, high-temperature heat pipe model, heat pipe heat exchanger model and energy conversion system model, etc. Using MATLAB/SIMULINK, the analysis platform suitable for this nuclear power system is built. Based on this platform, the steady state and transient performance of the system are characterized. In the steady state analysis, the simulated values of key nodes of the model are in general agreement with the calculated values of parameter matching, and the maximum relative error does not exceed 6 %. The effectiveness of characteristic analysis based on the built platform is verified. In the transient analysis, typical reactor system accidents and energy conversion system accident are simulated and analyzed respectively. The maximum temperature of the fuel assemblies in these accidents does not exceed 1550 K, which is lower than the selected material temperature safety limit. The inherent safety feature of the system is discussed.
{"title":"Characteristics analysis of the typical vehicle-mounted megawatt-scale heat pipe cooled reactor power system","authors":"Henglong Lin , Zeguang Li , Kan Wang , Weilin Zhuge , Ganglin Yu , Hongsheng Jiang , Jingkang Li , Jun Yang , Zilin Su , Huifu Wang","doi":"10.1016/j.anucene.2025.111467","DOIUrl":"10.1016/j.anucene.2025.111467","url":null,"abstract":"<div><div>Heat pipe cooled reactor (HPR) has the characteristics of small size and high safety, and can be used in special application scenarios such as vehicle-mounted mobility. In this paper, the Mobile Nuclear Power System based on HPR named MNPS-1000, which is designed to generate 1MWe, is selected for research and analysis. According to the design, a whole system model is established, which includes the reactor model, high-temperature heat pipe model, heat pipe heat exchanger model and energy conversion system model, etc. Using MATLAB/SIMULINK, the analysis platform suitable for this nuclear power system is built. Based on this platform, the steady state and transient performance of the system are characterized. In the steady state analysis, the simulated values of key nodes of the model are in general agreement with the calculated values of parameter matching, and the maximum relative error does not exceed 6 %. The effectiveness of characteristic analysis based on the built platform is verified. In the transient analysis, typical reactor system accidents and energy conversion system accident are simulated and analyzed respectively. The maximum temperature of the fuel assemblies in these accidents does not exceed 1550 K, which is lower than the selected material temperature safety limit. The inherent safety feature of the system is discussed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111467"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860309","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 : 2025-04-23DOI: 10.1016/j.anucene.2025.111450
Yufei Pan, Jimin Wang, Ruirui Xu, Xi Tao, Yue Zhang, Jie Ren, Yuan Tian, Zhigang Ge, Nengchuan Shu, Kang Xing, Xiaofei Wu
Neutron resonance data are significant importance for applications in nuclear engineering. Aimed to assess the discrepancies efficiently between the experimental measurements and evaluations, including ENDF/B-VIII.1, JEFF-3.3, JENDL-5, CENDL-3.2 and BROND-3.1. A resonance peak identification technique based on derivative method and Gaussian Filter is established to conduct systematic analysis on two resonance properties (integral values and central energy) of the experimental and evaluated data within the resolved resonance region. The deviations of the two properties of 238U are analyzed in this work. More than 1500 resonance peaks of neutron total and capture reactions are identified below neutron energy 20 keV. As a result, the evaluated central energies of (n, tot) resonances exhibit excellent agreement with experimental data, while those of (n, ) resonances show increasing deviation with rising energy levels; significant deviations of resonance integrals exist between experimental and evaluated data of (n, ), and (n, tot) demonstrate a better consistency. The current work efficiently identifies the energy regions that require improvement in the future.
{"title":"Neutron resonance integral analysis of 238U within resolved resonance region","authors":"Yufei Pan, Jimin Wang, Ruirui Xu, Xi Tao, Yue Zhang, Jie Ren, Yuan Tian, Zhigang Ge, Nengchuan Shu, Kang Xing, Xiaofei Wu","doi":"10.1016/j.anucene.2025.111450","DOIUrl":"10.1016/j.anucene.2025.111450","url":null,"abstract":"<div><div>Neutron resonance data are significant importance for applications in nuclear engineering. Aimed to assess the discrepancies efficiently between the experimental measurements and evaluations, including ENDF/B-VIII.1, JEFF-3.3, JENDL-5, CENDL-3.2 and BROND-3.1. A resonance peak identification technique based on derivative method and Gaussian Filter is established to conduct systematic analysis on two resonance properties (integral values and central energy) of the experimental and evaluated data within the resolved resonance region. The deviations of the two properties of <sup>238</sup>U are analyzed in this work. More than 1500 resonance peaks of neutron total and capture reactions are identified below neutron energy 20 keV. As a result, the evaluated central energies of (n, tot) resonances exhibit excellent agreement with experimental data, while those of (n, <span><math><mi>γ</mi></math></span>) resonances show increasing deviation with rising energy levels; significant deviations of resonance integrals exist between experimental and evaluated data of (n, <span><math><mi>γ</mi></math></span>), and (n, tot) demonstrate a better consistency. The current work efficiently identifies the energy regions that require improvement in the future.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111450"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860737","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 : 2025-04-23DOI: 10.1016/j.anucene.2025.111508
Xian Zhang , Shu Li , Xin Wang , Danhua ShangGuan , Shichang Liu
The Monte Carlo simulation of large-scale neutron transport problems has always faced the problem of slow computation. In order to fully exploit the acceleration advantage of heterogeneous parallelism on the Monte Carlo neutron transport simulation, this paper carries out research around the history-based neutron tracking algorithm, deeply explores the adaptation of the Monte Carlo algorithm and heterogeneous parallelism. Aiming at the thread divergence problem, optimization strategies for particle tracking algorithm are proposed to ensure load balancing among parallel threads. In addition, to mitigate the impact of global memory access latency, the memory layout of the particle state data is reasonably arranged by comprehensively considering the random memory access of Monte Carlo algorithm and the hardware characteristics of GPU. The reliability and efficiency of heterogeneous parallel algorithm are validated in calculations of benchmarks, the computing performance on an NVIDIA A800 GPU is equivalent to the performance of 62–87 CPU cores.
{"title":"Optimization of heterogeneous parallel algorithm for Monte Carlo neutron transport simulation aiming at thread divergence Issues","authors":"Xian Zhang , Shu Li , Xin Wang , Danhua ShangGuan , Shichang Liu","doi":"10.1016/j.anucene.2025.111508","DOIUrl":"10.1016/j.anucene.2025.111508","url":null,"abstract":"<div><div>The Monte Carlo simulation of large-scale neutron transport problems has always faced the problem of slow computation. In order to fully exploit the acceleration advantage of heterogeneous parallelism on the Monte Carlo neutron transport simulation, this paper carries out research around the history-based neutron tracking algorithm, deeply explores the adaptation of the Monte Carlo algorithm and heterogeneous parallelism. Aiming at the thread divergence problem, optimization strategies for particle tracking algorithm are proposed to ensure load balancing among parallel threads. In addition, to mitigate the impact of global memory access latency, the memory layout of the particle state data is reasonably arranged by comprehensively considering the random memory access of Monte Carlo algorithm and the hardware characteristics of GPU. The reliability and efficiency of heterogeneous parallel algorithm are validated in calculations of benchmarks, the computing performance on an NVIDIA A800 GPU is equivalent to the performance of 62–87 CPU cores.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111508"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863598","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 : 2025-04-23DOI: 10.1016/j.anucene.2025.111511
A. Ravi Kiran , M.K. Agrawal , Renuka Patel , S.K. Sinha
A tray rod assembly in a research reactor is a facility that irradiates samples inside reactor core. For on-power lifting of the tray rod assembly, it is necessary to ensure its integrity under axial flow induced vibrations, resulting from coolant flow. In the present work, methodology for qualifying the axial flow induced vibration response of a tray rod assembly is demonstrated. Experiments are conducted to study the axial flow induced vibration response of an ‘on-power tray rod’ of a typical Indian Research Reactor (IRR), using an in-house flow test facility. Three conditions are proposed for the qualification of ‘on-power’ tray rod assembly under axial flow induced vibration load. First one is the permissible vibration velocity, second one is the permissible displacement and the third one is the permissible flow velocity. Free vibration analysis of the assembly is carried out to obtain the dynamic characteristics and to compare with experimental results.
{"title":"Experimental study of internal flow induced vibration of tray rod assembly during on-power handling in a typical Indian research reactor","authors":"A. Ravi Kiran , M.K. Agrawal , Renuka Patel , S.K. Sinha","doi":"10.1016/j.anucene.2025.111511","DOIUrl":"10.1016/j.anucene.2025.111511","url":null,"abstract":"<div><div>A tray rod assembly in a research reactor is a facility that irradiates samples inside reactor core. For on-power lifting of the tray rod assembly, it is necessary to ensure its integrity under axial flow induced vibrations, resulting from coolant flow. In the present work, methodology for qualifying the axial flow induced vibration response of a tray rod assembly is demonstrated. Experiments are conducted to study the axial flow induced vibration response of an ‘on-power tray rod’ of a typical Indian Research Reactor (IRR), using an in-house flow test facility. Three conditions are proposed for the qualification of ‘on-power’ tray rod assembly under axial flow induced vibration load. First one is the permissible vibration velocity, second one is the permissible displacement and the third one is the permissible flow velocity. Free vibration analysis of the assembly is carried out to obtain the dynamic characteristics and to compare with experimental results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111511"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860727","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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