Pub Date : 2026-01-06DOI: 10.1016/j.anucene.2025.112060
Peter J. Kriemadis, Adriaan Buijs
The Zero Energy Deuterium (ZED-2) reactor is a zero-power research reactor located at the Chalk River site of Canadian Nuclear Laboratories (CNL). The reactor was built to assist in neutronics code validation efforts for CANadian Deuterium Uranium (CANDU) reactors, but may find further use in the validation of computer codes used in the design of Small Modular Reactors (SMRs). This paper describes the application of the OpenMC and SERPENT 2 codes to two published benchmarks for ZED-2 neutronics experiments. The results were then compared to MCNP and MONK code results on file. Experiments from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook were reviewed to establish the differences one might expect from Monte Carlo code-to-code comparisons. The completed benchmarks were assessed against this review. In this manner, the OpenMC code is validated both against an experiment and against other validated codes.
{"title":"ZED-2 benchmarks performed in OpenMC and Serpent 2: A validation exercise for OpenMC applications","authors":"Peter J. Kriemadis, Adriaan Buijs","doi":"10.1016/j.anucene.2025.112060","DOIUrl":"10.1016/j.anucene.2025.112060","url":null,"abstract":"<div><div>The Zero Energy Deuterium (ZED-2) reactor is a zero-power research reactor located at the Chalk River site of Canadian Nuclear Laboratories (CNL). The reactor was built to assist in neutronics code validation efforts for CANadian Deuterium Uranium (CANDU) reactors, but may find further use in the validation of computer codes used in the design of Small Modular Reactors (SMRs). This paper describes the application of the OpenMC and SERPENT<!--> <!-->2 codes to two published benchmarks for ZED-2 neutronics experiments. The results were then compared to MCNP and MONK code results on file. Experiments from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook were reviewed to establish the differences one might expect from Monte Carlo code-to-code comparisons. The completed benchmarks were assessed against this review. In this manner, the OpenMC code is validated both against an experiment and against other validated codes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112060"},"PeriodicalIF":2.3,"publicationDate":"2026-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922117","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 : 2026-01-03DOI: 10.1016/j.anucene.2025.112104
Jing Zhang, Shurong Ding
FCM fuel, composed of TRISO particles and SiC matrix, is a typical accident tolerant fuel that holds a promising application prospect in various advanced nuclear reactors. Optimization of the microstructure of TRISO (Tri-Structural Isotropic) particles is crucial for enhancing both the safety and economy of nuclear fuel systems. In this study, the recently published novel fission gas swelling model or volume-growth strain models for the fuel kernel and the buffer layer are involved, enabling more accurate simulation of the irradiation thermo-mechanical coupling behaviors of FCM fuel. The three-dimensional mechanical constitutive relations, stress update algorithms and consistent stiffness moduli for the points within the buffer layer and PyC layer are newly formulated, and the corresponding procedures are developed. With the fuel kernel volume fraction and SiC layer dimensions kept constant, the effects of the thicknesses of the buffer layer, IPyC layer, and OPyC layer on the safety of FCM fuel are investigated. The research findings indicate that: (1) Increasing the buffer layer thickness can effectively improve its ability to accommodate kernel swelling, thereby markedly weakening the mechanical interactions between different parts; (2) With an increase of the buffer layer thickness from 50 µm to 80 µm, the peak first principal stresses in the SiC layer and the matrix decrease by 51 % and 79 %, respectively, leading to a significantly reduced failure risk; (3) A strategic redistribution of layer thicknesses can significantly strengthen the TRISO fuel safety, particularly by increasing the buffer layer thickness while decreasing both inner and outer dense PyC layer thicknesses, without altering other microstructural parameters. This study can provide theoretical guidance and analytical tools for the advanced manufacturing and optimization design of FCM fuel.
{"title":"Effects of coating layer thicknesses on the thermo-mechanical coupling behaviors of FCM fuels","authors":"Jing Zhang, Shurong Ding","doi":"10.1016/j.anucene.2025.112104","DOIUrl":"10.1016/j.anucene.2025.112104","url":null,"abstract":"<div><div>FCM fuel, composed of TRISO particles and SiC matrix, is a typical accident tolerant fuel that holds a promising application prospect in various advanced nuclear reactors. Optimization of the microstructure of TRISO (Tri-Structural Isotropic) particles is crucial for enhancing both the safety and economy of nuclear fuel systems. In this study, the recently published novel fission gas swelling model or volume-growth strain models for the fuel kernel and the buffer layer are involved, enabling more accurate simulation of the irradiation thermo-mechanical coupling behaviors of FCM fuel. The three-dimensional mechanical constitutive relations, stress update algorithms and consistent stiffness moduli for the points within the buffer layer and PyC layer are newly formulated, and the corresponding procedures are developed. With the fuel kernel volume fraction and SiC layer dimensions kept constant, the effects of the thicknesses of the buffer layer, IPyC layer, and OPyC layer on the safety of FCM fuel are investigated. The research findings indicate that: (1) Increasing the buffer layer thickness can effectively improve its ability to accommodate kernel swelling, thereby markedly weakening the mechanical interactions between different parts; (2) With an increase of the buffer layer thickness from 50 µm to 80 µm, the peak first principal stresses in the SiC layer and the matrix decrease by 51 % and 79 %, respectively, leading to a significantly reduced failure risk; (3) A strategic redistribution of layer thicknesses can significantly strengthen the TRISO fuel safety, particularly by increasing the buffer layer thickness while decreasing both inner and outer dense PyC layer thicknesses, without altering other microstructural parameters. This study can provide theoretical guidance and analytical tools for the advanced manufacturing and optimization design of FCM fuel.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112104"},"PeriodicalIF":2.3,"publicationDate":"2026-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145881360","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 : 2026-01-03DOI: 10.1016/j.anucene.2025.112101
Jiaxing Ren , Quanbo Li , Gongyao Liu , Weiqiang Xu , Ruifeng Tian , Puzhen Gao , Shouxu Qiao , Sichao Tan
Due to the complex effects of advection, pressure drop, and bubble interactions, the two-phase flow parameters change dynamically with flow development. This paper conducted an experimental study on air–water two-phase flow in a 5 × 5 rod bundle using a four-sensor conductivity probe and differential pressure transducers. An interfacial area parameter database containing over 5000 measurement points is established from the entire cross-sectional measurement at five axial positions. The enhanced bubble coalescence efficiency due to the geometry of subchannels results in an increasing trend of bubble velocity and a decreasing trend of interfacial area concentration. Cap-bubbly flows are found to have larger relative velocities than bubbly flows through the one-dimensional drift-flux analysis. The two-phase frictional pressure drop is calculated by the measured void fraction with the probe and used to evaluate the existing prediction models. The coupling relationship between the void fraction and pressure drop is also analyzed through operating flow parameters.
{"title":"Research on the characteristics of interfacial area transport and flow resistance for upward two-phase flow in a 5 × 5 rod bundle","authors":"Jiaxing Ren , Quanbo Li , Gongyao Liu , Weiqiang Xu , Ruifeng Tian , Puzhen Gao , Shouxu Qiao , Sichao Tan","doi":"10.1016/j.anucene.2025.112101","DOIUrl":"10.1016/j.anucene.2025.112101","url":null,"abstract":"<div><div>Due to the complex effects of advection, pressure drop, and bubble interactions, the two-phase flow parameters change dynamically with flow development. This paper conducted an experimental study on air–water two-phase flow in a 5 × 5 rod bundle using a four-sensor conductivity probe and differential pressure transducers. An interfacial area parameter database containing over 5000 measurement points is established from the entire cross-sectional measurement at five axial positions. The enhanced bubble coalescence efficiency due to the geometry of subchannels results in an increasing trend of bubble velocity and a decreasing trend of interfacial area concentration. Cap-bubbly flows are found to have larger relative velocities than bubbly flows through the one-dimensional drift-flux analysis. The two-phase frictional pressure drop is calculated by the measured void fraction with the probe and used to evaluate the existing prediction models. The coupling relationship between the void fraction and pressure drop is also analyzed through operating flow parameters.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112101"},"PeriodicalIF":2.3,"publicationDate":"2026-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145881355","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 : 2026-01-02DOI: 10.1016/j.anucene.2025.112100
Ying Guan , Yang Li , Hualei Jiang , Daqing Wang , Daochuan Ge , Huaping Mei , Gui Fang , Lifu Gao , Haixia Wang
Bubble migration during a steam generator tube rupture (SGTR) accident in lead-cooled fast reactor (LFR) has garnered significant attention. This phenomenon can cause localized heat transfer deterioration and power fluctuations, posing substantial safety risks to the reactor. This paper presents a novel detection approach integrated with dynamic tracking for multiple micro bubbles, specifically designed for transparent liquid similarity experiments in SGTR research. The proposed method, named DAM-YOLO, incorporates dual attention mechanisms, a lightweight upsampling operator, and content-aware reassembly of features to enhance feature extraction capability and feature fusion performance for micro bubbles. Furthermore, by adopting the DeepSORT algorithm combined with the complete intersection over union (CIoU) matching metric, the issue of multiple target loss during tracking process is effectively addressed. In this study, bubble datasets were acquired from a self-developed similarity experimental facility. The results demonstrate that the precision (P), mean average precision (mAP), multiple object tracking accuracy (MOTA), multiple object tracking precision (MOTP), and id f1 score (IDF1) of the proposed model reach 96.4%, 95.6%, 85.27%, 86.99%, and 92.63%, respectively. This research can provide efficient intelligent technical support for analyzing the migration process of multiple micro bubbles in fluid dynamics studies.
{"title":"Computer vision-based detection and dynamic tracking of multiple micro bubbles in transparent liquid similarity experiments for SGTR in lead-cooled fast reactor","authors":"Ying Guan , Yang Li , Hualei Jiang , Daqing Wang , Daochuan Ge , Huaping Mei , Gui Fang , Lifu Gao , Haixia Wang","doi":"10.1016/j.anucene.2025.112100","DOIUrl":"10.1016/j.anucene.2025.112100","url":null,"abstract":"<div><div>Bubble migration during a steam generator tube rupture (SGTR) accident in lead-cooled fast reactor (LFR) has garnered significant attention. This phenomenon can cause localized heat transfer deterioration and power fluctuations, posing substantial safety risks to the reactor. This paper presents<!--> <!-->a novel detection approach integrated with dynamic tracking for multiple micro bubbles, specifically designed for transparent liquid similarity experiments in SGTR research. The proposed method, named <span>DAM</span>-YOLO, incorporates dual attention mechanisms, a lightweight upsampling operator, and content-aware reassembly of features to enhance feature extraction capability and feature fusion performance for micro bubbles. <span>Furthermore</span>, by adopting the DeepSORT algorithm combined with the complete intersection over union (CIoU) matching metric, the issue of multiple target loss during tracking process is effectively addressed. In this study, bubble datasets were acquired<!--> <!-->from a self-developed similarity experimental facility. The results demonstrate that the precision (P), mean average precision (mAP), multiple object tracking accuracy (MOTA), multiple object tracking precision (MOTP), and id f1 score (IDF1) of the proposed model reach 96.4%, 95.6%, 85.27%, 86.99%, and 92.63%, respectively. This research can provide efficient intelligent technical support for analyzing the migration process of multiple micro bubbles in fluid dynamics studies.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112100"},"PeriodicalIF":2.3,"publicationDate":"2026-01-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145881354","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 : 2026-01-02DOI: 10.1016/j.anucene.2025.112106
Tomáš Peltan , Tomáš Czakoj , Vlastimil Juříček , Jan Šimon , Michal Košťál
This study of a slightly moderated neutron reference field shaped by graphite has been established at the LR-0 reactor. This experimental configuration modifies the reference neutron spectrum toward lower energies while preserving spatial homogeneity using graphite. This core configuration was characterized through an extensive neutron flux mapping by four independent irradiation campaigns utilizing activation foil detectors. High-purity and well-known dosimetry reactions using activation foils of Au, Cu, Mn, Ta, and Ni were strategically positioned throughout the graphite insertion for neutron shape mapping. The resulting reaction rates were derived from gamma spectrometry using a well-defined HPGe detector. The measured data provide detailed spatial and spectral information on the neutron flux distribution and confirm the reproducibility and stability of the investigated volume in the graphite field, which can be validated as a new neutron reference field. This field can be used to research in advanced reactor systems and IV. gen reactors.
{"title":"A development of a new slightly moderated reference field in graphite insertion in the LR-0 reactor","authors":"Tomáš Peltan , Tomáš Czakoj , Vlastimil Juříček , Jan Šimon , Michal Košťál","doi":"10.1016/j.anucene.2025.112106","DOIUrl":"10.1016/j.anucene.2025.112106","url":null,"abstract":"<div><div>This study of a slightly moderated neutron reference field shaped by graphite has been established at the LR-0 reactor. This experimental configuration modifies the reference neutron spectrum toward lower energies while preserving spatial homogeneity using graphite. This core configuration was characterized through an extensive neutron flux mapping by four independent irradiation campaigns utilizing activation foil detectors. High-purity and well-known dosimetry reactions using activation foils of Au, Cu, Mn, Ta, and Ni were strategically positioned throughout the graphite insertion for neutron shape mapping. The resulting reaction rates were derived from gamma spectrometry using a well-defined HPGe detector. The measured data provide detailed spatial and spectral information on the neutron flux distribution and confirm the reproducibility and stability of the investigated volume in the graphite field, which can be validated as a new neutron reference field. This field can be used to research in advanced reactor systems and IV. gen reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112106"},"PeriodicalIF":2.3,"publicationDate":"2026-01-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145881356","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 : 2025-12-31DOI: 10.1016/j.anucene.2025.112109
Ji Liu , Yongkuo Liu , Zhouxin Shi , Jiarong Gao , Yukun Liu , Zhen Wang , Guohua Wu
Due to the unique nature of nuclear power plants, highly reliable fault diagnosis methods are required to ensure operational safety and stability. To fully capture the spatiotemporal dependencies in multivariate time series (MTS) data and improve the accuracy of fault diagnosis in nuclear power plant systems, this paper proposes a hybrid diagnostic framework integrating Crossformer and Support Vector Machine (SVM), referred to as the Crossformer-SVM model. First, using the PCTRAN simulation platform and the Fuqing simulation machine as data sources, fault datasets with noise and without noise are constructed. Then, the Crossformer model is employed to hierarchically extract the spatiotemporal features of the system fault data, which are used as inputs for the SVM classifier. Finally, the SVM classifier is used to identify the fault modes of the system. In addition, a comparative experiment is conducted between the proposed Crossformer-SVM model and other deep learning models, such as CNN-LSTM. The experimental results show that, compared to other deep learning fault diagnosis models, the proposed method achieves the highest accuracy, with a minimum accuracy of 99.20% for the two types of noise-free datasets. It also maintains excellent diagnostic performance under noise, with diagnostic accuracies of 98.92% and 98.88% for the Fuqing simulator and PCTRAN data, respectively. This provides a reliable fault diagnosis method for nuclear power plant systems.
{"title":"Research on fault diagnosis method for nuclear power plants based on crossformer-SVM","authors":"Ji Liu , Yongkuo Liu , Zhouxin Shi , Jiarong Gao , Yukun Liu , Zhen Wang , Guohua Wu","doi":"10.1016/j.anucene.2025.112109","DOIUrl":"10.1016/j.anucene.2025.112109","url":null,"abstract":"<div><div>Due to the unique nature of nuclear power plants, highly reliable fault diagnosis methods are required to ensure operational safety and stability. To fully capture the spatiotemporal dependencies in multivariate time series (MTS) data and improve the accuracy of fault diagnosis in nuclear power plant systems, this paper proposes a hybrid diagnostic framework integrating Crossformer and Support Vector Machine (SVM), referred to as the Crossformer-SVM model. First, using the PCTRAN simulation platform and the Fuqing simulation machine as data sources, fault datasets with noise and without noise are constructed. Then, the Crossformer model is employed to hierarchically extract the spatiotemporal features of the system fault data, which are used as inputs for the SVM classifier. Finally, the SVM classifier is used to identify the fault modes of the system. In addition, a comparative experiment is conducted between the proposed Crossformer-SVM model and other deep learning models, such as CNN-LSTM. The experimental results show that, compared to other deep learning fault diagnosis models, the proposed method achieves the highest accuracy, with a minimum accuracy of 99.20% for the two types of noise-free datasets. It also maintains excellent diagnostic performance under noise, with diagnostic accuracies of 98.92% and 98.88% for the Fuqing simulator and PCTRAN data, respectively. This provides a reliable fault diagnosis method for nuclear power plant systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112109"},"PeriodicalIF":2.3,"publicationDate":"2025-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145881415","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 : 2025-12-31DOI: 10.1016/j.anucene.2025.112088
Siyuan Wu, Jinpeng He, Weiguo Gu, Deyi Chen, Baojie Nie, Dezhong Wang
The dispersion uplift of the aerosol plume due to the complex mountain subsurface and the residence of the plume in the clockwise turnaround flow behind the hill are still a major obstacle in accurately predicting the dispersion concentration distributions of aerosols in the atmospheric emissions from nuclear power plants, despite the use of complex numerical models. In particular, the effect of hill combination on dispersion trajectories and concentration distributions has not received widely attention in previous experimental studies. The focus of this study is on the effect of the distance between two identical parallel aligned hills on the dispersion trajectory and concentration distribution of the aerosol plume under neutral atmospheric conditions. A combination of planar particle laser concentration measurement method and CFD simulation was used to characterize the flow reattachment and plume trajectories as well as the dispersion distribution. The results show that when the interval between the two hills is less than , the total reattachment length remains at as in the case of a single hill, and the dispersion trajectory is basically consistent with that of a single hill. The concentration accumulates on the windward side of the second hill between valleys. When the distance between the two hills is greater than , the reattachment length of the first hill recovers to as in the case of a single hill, and the second hill begins to move away from the influence area of the first hill.
{"title":"Effects of two-dimensional hills combination with distance variation on the aerosols dispersion based on wind tunnel experiment","authors":"Siyuan Wu, Jinpeng He, Weiguo Gu, Deyi Chen, Baojie Nie, Dezhong Wang","doi":"10.1016/j.anucene.2025.112088","DOIUrl":"10.1016/j.anucene.2025.112088","url":null,"abstract":"<div><div>The dispersion uplift of the aerosol plume due to the complex mountain subsurface and the residence of the plume in the clockwise turnaround flow behind the hill are still a major obstacle in accurately predicting the dispersion concentration distributions of aerosols in the atmospheric emissions from nuclear power plants, despite the use of complex numerical models. In particular, the effect of hill combination on dispersion trajectories and concentration distributions has not received widely attention in previous experimental studies. The focus of this study is on the effect of the distance between two identical parallel aligned hills on the dispersion trajectory and concentration distribution of the aerosol plume under neutral atmospheric conditions. A combination of planar particle laser concentration measurement method and CFD simulation was used to characterize the flow reattachment and plume trajectories as well as the dispersion distribution. The results show that when the interval between the two hills is less than <span><math><mrow><mn>4</mn><mi>H</mi></mrow></math></span>, the total reattachment length remains at <span><math><mrow><mn>9</mn><mi>H</mi></mrow></math></span> as in the case of a single hill, and the dispersion trajectory is basically consistent with that of a single hill. The concentration accumulates on the windward side of the second hill between valleys. When the distance between the two hills is greater than <span><math><mrow><mn>12</mn><mi>H</mi></mrow></math></span>, the reattachment length of the first hill recovers to <span><math><mrow><mn>9</mn><mi>H</mi></mrow></math></span> as in the case of a single hill, and the second hill begins to move away from the influence area of the first hill.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112088"},"PeriodicalIF":2.3,"publicationDate":"2025-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145881358","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 : 2025-12-31DOI: 10.1016/j.anucene.2025.112068
Mahmoud Eltawila , Pierre-Clément A. Simon , Guillaume L. Giudicelli , Helen Brooks , Nicholas Wozniak , April J. Novak
Geometry deformation due to thermal expansion influences neutron transport in many systems. Studying this phenomenon involves coupling models for neutronics, thermal hydraulics, and solid mechanics. To enable high fidelity modeling of these coupled physics, new capabilities were introduced in Cardinal, coupling OpenMC Monte Carlo particle transport models with MOOSE thermomechanical physics on unstructured moving-mesh geometries. In this work, we present a fully open-source capability leveraging on-the-fly mesh skinning to automatically regenerate OpenMC geometry, which allows multiphysics feedback from temperature, density, and geometry changes. The new capability is verified using an analytic benchmark slab problem, which couples S neutron transport with thermal conduction, convective boundary conditions, Doppler-broadened cross sections, and nonlinear thermal expansion effects along the heated slab. Cardinal reproduces the analytic solutions for the neutron flux, heating, k, and temperature with demonstrated convergence in various error terms including mesh resolution and cross section temperature library spacing. For the nominal benchmark conditions and with a fine mesh, maximum relative errors for neutron flux, temperature, and heating are lower than 1%, while errors in integral quantities such as and slab length are within 1 pcm and 48 µm, respectively. This work (i) presents a new numerical approach to thermomechanics coupling with OpenMC models, (ii) is the first (to our knowledge) to utilize a mechanical partial differential equation (PDE) solution to solve the (Griesheimer and Kooreman, 2022) analytic benchmark, and (iii) develops this verified capability within an open-source package.
{"title":"Thermomechanics coupling to Monte Carlo particle transport on unstructured mesh geometries using Cardinal","authors":"Mahmoud Eltawila , Pierre-Clément A. Simon , Guillaume L. Giudicelli , Helen Brooks , Nicholas Wozniak , April J. Novak","doi":"10.1016/j.anucene.2025.112068","DOIUrl":"10.1016/j.anucene.2025.112068","url":null,"abstract":"<div><div>Geometry deformation due to thermal expansion influences neutron transport in many systems. Studying this phenomenon involves coupling models for neutronics, thermal hydraulics, and solid mechanics. To enable high fidelity modeling of these coupled physics, new capabilities were introduced in Cardinal, coupling OpenMC Monte Carlo particle transport models with MOOSE thermomechanical physics on unstructured moving-mesh geometries. In this work, we present a fully open-source capability leveraging on-the-fly mesh skinning to automatically regenerate OpenMC geometry, which allows multiphysics feedback from temperature, density, and geometry changes. The new capability is verified using an analytic benchmark slab problem, which couples S<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span> neutron transport with thermal conduction, convective boundary conditions, Doppler-broadened cross sections, and nonlinear thermal expansion effects along the heated slab. Cardinal reproduces the analytic solutions for the neutron flux, heating, k<span><math><msub><mrow></mrow><mrow><mtext>eff</mtext></mrow></msub></math></span>, and temperature with demonstrated convergence in various error terms including mesh resolution and cross section temperature library spacing. For the nominal benchmark conditions and with a fine mesh, maximum relative errors for neutron flux, temperature, and heating are lower than 1%, while errors in integral quantities such as <span><math><msub><mrow><mi>k</mi></mrow><mrow><mtext>eff</mtext></mrow></msub></math></span> and slab length are within 1 pcm and 48 µm, respectively. This work (i) presents a new numerical approach to thermomechanics coupling with OpenMC models, (ii) is the first (to our knowledge) to utilize a mechanical partial differential equation (PDE) solution to solve the (Griesheimer and Kooreman, 2022) analytic benchmark, and (iii) develops this verified capability within an open-source package.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112068"},"PeriodicalIF":2.3,"publicationDate":"2025-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145881414","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 : 2025-12-31DOI: 10.1016/j.anucene.2025.112102
Tiandi Fan , Huajie Wu , Changzheng Chen , Xianming Sun , lipeng Wang
During the transportation of large-scale equipment, road excitation can trigger resonance, potentially impairing the structural stability of the vehicle system. Moreover, due to its special structure and non-uniform mass distribution—such as reactor pool mass concentration from high-density Lead-Bismuth Eutectic (LBE) coolant—transportation vibration response and stability have become urgent key technical issues. Besides, liquid LBE coolant may slosh during transportation, requiring stricter stability rules. However, current research focuses more on anti-vibration design or isolation, with few studies on the equipment’s vibration. To solve this problem, this study uses Lagrange’s differential equations to create a new separated vibration model. It explains how mobile LBE reactors respond to vibration, and analyzes vibration responses under different road classes, driving speeds, and loading masses. This study confirms the developed separated model’s ability to capture on-board equipment and vehicle body vibration responses, offering references for large equipment transportation safety and insights for mobile LBE reactor design.
{"title":"Vibration analysis of large equipment with Non-Uniform mass under road transportation","authors":"Tiandi Fan , Huajie Wu , Changzheng Chen , Xianming Sun , lipeng Wang","doi":"10.1016/j.anucene.2025.112102","DOIUrl":"10.1016/j.anucene.2025.112102","url":null,"abstract":"<div><div>During the transportation of large-scale equipment, road excitation can trigger resonance, potentially impairing the structural stability of the vehicle system. Moreover, due to its special structure and non-uniform mass distribution—such as reactor pool mass concentration from high-density Lead-Bismuth Eutectic (LBE) coolant—transportation vibration response and stability have become urgent key technical issues. Besides, liquid LBE coolant may slosh during transportation, requiring stricter stability rules. However, current research focuses more on anti-vibration design or isolation, with few studies on the equipment’s vibration. To solve this problem, this study uses Lagrange’s differential equations to create a new separated vibration model. It explains how mobile LBE reactors respond to vibration, and analyzes vibration responses under different road classes, driving speeds, and loading masses. This study confirms the developed separated model’s ability to capture on-board equipment and vehicle body vibration responses, offering references for large equipment transportation safety and insights for mobile LBE reactor design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"229 ","pages":"Article 112102"},"PeriodicalIF":2.3,"publicationDate":"2025-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145881357","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 : 2025-12-30DOI: 10.1016/j.anucene.2025.112108
Arya Pramana Sembiring , Juyoul Kim
Gerrit Agustinus Siwabessy Multi-Purpose Reactor (RSG-GAS) is Indonesia’s largest research reactor. This research reactor needs radiation risk mitigation in a densely populated area. This study examines the radiation shielding effectiveness of Building No. 93 as an offsite emergency center during a Beyond Design-Basis Accident (BDBA). An Anticipated Transient Without Scram (ATWS) scenario with five fuel element meltdowns was simulated using HotSpot, a Gaussian plume-based atmospheric dispersion software, and dose attenuation was performed using Monte Carlo. The result shows that the dual-layered concrete wall can reduce the incoming total dose rate to 0.558 mSv/hour, allowing safe occupancy for 3.74 days before reaching Indonesia’s emergency dose limit of 50 mSv for workers and the public. This study confirms that Building No. 93 meets regulatory standards and is suitable as an offsite emergency center for emergency operators during severe nuclear incidents. This study established a framework for future emergency evaluation and management strategy.
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