Pub Date : 2026-01-16DOI: 10.1016/j.anucene.2026.112116
Romain Henry, Jérémy Bousquet, Armin Seubert
This paper presents a new feature of the Finite Element Method (FEM) code FENNECS (Finite ElemeNt NEutroniCS) for modelling reactivity control systems. The interface between materials within a finite element (usually referred to as a mixed element) is modelled using a flux-weighting method. While the method has demonstrated its accuracy in modelling the vertical movement of control rods in traditional Light Water and Fast Reactors (LWR and FR), it has limitations in modelling the rotation of control drums.
The projection-based cusping treatment is another method that defines an effective homogenized cross-section for the mixed element. Unlike the flux-weighting method, this method does not involve any approximations. Instead, it exactly solves the weak form of the neutron diffusion equation.
In order to illustrate the appropriate implementation of the method in the code, three exercises were solved. A comparison with the legacy flux-weighting model was performed, highlighting the benefits of the projection-based de-cusping method.
In every case, if it is not completely removed, the cusping effect is mitigated, enabling the production of a solution compatible with nuclear safety analysis. Furthermore, it has been demonstrated that the projection-based method clearly outperforms the flux and volume weighting method in terms of accuracy.
In terms of runtime, the projection-based method has demonstrated an average reduction of 40% for control rod exercises, while control drum exercises have shown a reduction of 15%.
本文介绍了用于反应性控制系统建模的有限元方法（FEM）代码FENNECS （Finite Element NEutroniCS）的一个新特性。有限单元（通常称为混合单元）内材料之间的界面采用通量加权法建模。虽然该方法在模拟传统轻水快堆（LWR和FR）中控制棒的垂直运动方面已经证明了它的准确性，但它在模拟控制鼓的旋转方面存在局限性。基于投影的尖化处理是另一种定义混合单元有效均匀截面的方法。与通量加权法不同，该方法不涉及任何近似。相反，它精确地解出了中子扩散方程的弱形式。为了说明代码中方法的适当实现，解决了三个练习。与传统的通量加权模型进行了比较，突出了基于投影的去尖化方法的优点。在任何情况下，如果没有完全清除，则可以减轻尖刺效应，从而能够产生与核安全分析相容的解决方案。此外，基于投影的方法在精度方面明显优于通量和体积加权方法。在运行时间方面，基于投影的方法表明，控制棒练习平均减少40%，而控制鼓练习平均减少15%。
{"title":"Implementation of a projection-based control rod de-cusping method in the Finite Element Neutronic Code FENNECS","authors":"Romain Henry, Jérémy Bousquet, Armin Seubert","doi":"10.1016/j.anucene.2026.112116","DOIUrl":"10.1016/j.anucene.2026.112116","url":null,"abstract":"<div><div>This paper presents a new feature of the Finite Element Method (FEM) code FENNECS (Finite ElemeNt NEutroniCS) for modelling reactivity control systems. The interface between materials within a finite element (usually referred to as a mixed element) is modelled using a flux-weighting method. While the method has demonstrated its accuracy in modelling the vertical movement of control rods in traditional Light Water and Fast Reactors (LWR and FR), it has limitations in modelling the rotation of control drums.</div><div>The projection-based cusping treatment is another method that defines an effective homogenized cross-section for the mixed element. Unlike the flux-weighting method, this method does not involve any approximations. Instead, it exactly solves the weak form of the neutron diffusion equation.</div><div>In order to illustrate the appropriate implementation of the method in the code, three exercises were solved. A comparison with the legacy flux-weighting model was performed, highlighting the benefits of the projection-based de-cusping method.</div><div>In every case, if it is not completely removed, the cusping effect is mitigated, enabling the production of a solution compatible with nuclear safety analysis. Furthermore, it has been demonstrated that the projection-based method clearly outperforms the flux and volume weighting method in terms of accuracy.</div><div>In terms of runtime, the projection-based method has demonstrated an average reduction of 40% for control rod exercises, while control drum exercises have shown a reduction of 15%.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112116"},"PeriodicalIF":2.3,"publicationDate":"2026-01-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145975288","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 : 2026-01-14DOI: 10.1016/j.anucene.2025.112107
Cristian G. de Oliveira , Alirio J.S. Piña , Antonella L. Costa , Claubia Pereira , Clarysson A.M. Silva , Damian E. Ramajo , Dario M. Godino
Small Modular Reactors (SMRs) are increasingly pursued to provide reliable low-carbon power, with Pressurized Water Reactors (PWRs) offering particular advantages due to extensive operational experience, well-characterized coolant properties, and an established industrial infrastructure. In this context, the present study develops a detailed conjugate heat-transfer (CHT) model of the hottest fuel assembly in a PWR-type SMR using the open-source CFD toolbox OpenFOAM. Exploiting geometric symmetry, the fuel and coolant regions were resolved with finite volumes, while the thin helium gap and cladding layers were modeled as lumped thermal resistances. A four-case mesh sensitivity study confirmed mesh-independent predictions of the average outlet coolant temperature and quantified the influence of spatial refinement on hydraulic-loss estimation. Steady-state CHT simulations provided axial and radial temperature distributions in both the fuel and coolant, together with detailed velocity and pressure fields. The results captured key physical features, including spacer-grid-induced velocity reductions and the coupling between coolant heating, density decrease, and local flow acceleration. The predicted average outlet coolant temperature (607 K) was consistent with reference operating conditions, while the total pressure drop across the assembly was found to be on the order of 20 kPa, reflecting the combined effect of frictional losses and spacer-grid resistance. A transient Loss-of-Flow Accident (LOFA) was also simulated by imposing a 50% reduction in the inlet mass flow rate. The analysis revealed oscillations in the outlet temperature driven by variable-density effects, with an average increase of approximately 15 K relative to steady-state conditions. Fuel centerline temperatures increased accordingly but remained safely below the UO melting point. Overall, the proposed methodology demonstrates the capability of open-source CFD tools to predict coupled neutronic–thermal–hydraulic behavior at the fuel-assembly scale in PWR-type SMRs, providing a solid foundation for future extensions to multiphase modeling and beyond-design-basis accident scenarios.
{"title":"Development of a local power peaking analysis methodology using OpenFOAM for a pressurized water-cooled small modular reactor","authors":"Cristian G. de Oliveira , Alirio J.S. Piña , Antonella L. Costa , Claubia Pereira , Clarysson A.M. Silva , Damian E. Ramajo , Dario M. Godino","doi":"10.1016/j.anucene.2025.112107","DOIUrl":"10.1016/j.anucene.2025.112107","url":null,"abstract":"<div><div>Small Modular Reactors (SMRs) are increasingly pursued to provide reliable low-carbon power, with Pressurized Water Reactors (PWRs) offering particular advantages due to extensive operational experience, well-characterized coolant properties, and an established industrial infrastructure. In this context, the present study develops a detailed conjugate heat-transfer (CHT) model of the hottest fuel assembly in a PWR-type SMR using the open-source CFD toolbox OpenFOAM. Exploiting geometric symmetry, the fuel and coolant regions were resolved with finite volumes, while the thin helium gap and cladding layers were modeled as lumped thermal resistances. A four-case mesh sensitivity study confirmed mesh-independent predictions of the average outlet coolant temperature and quantified the influence of spatial refinement on hydraulic-loss estimation. Steady-state CHT simulations provided axial and radial temperature distributions in both the fuel and coolant, together with detailed velocity and pressure fields. The results captured key physical features, including spacer-grid-induced velocity reductions and the coupling between coolant heating, density decrease, and local flow acceleration. The predicted average outlet coolant temperature (607 K) was consistent with reference operating conditions, while the total pressure drop across the assembly was found to be on the order of 20 kPa, reflecting the combined effect of frictional losses and spacer-grid resistance. A transient Loss-of-Flow Accident (LOFA) was also simulated by imposing a 50% reduction in the inlet mass flow rate. The analysis revealed oscillations in the outlet temperature driven by variable-density effects, with an average increase of approximately 15 K relative to steady-state conditions. Fuel centerline temperatures increased accordingly but remained safely below the UO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span> melting point. Overall, the proposed methodology demonstrates the capability of open-source CFD tools to predict coupled neutronic–thermal–hydraulic behavior at the fuel-assembly scale in PWR-type SMRs, providing a solid foundation for future extensions to multiphase modeling and beyond-design-basis accident scenarios.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112107"},"PeriodicalIF":2.3,"publicationDate":"2026-01-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145975067","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 : 2026-01-14DOI: 10.1016/j.anucene.2026.112139
Zaid Abulawi, Doyeong Lim, Abhiram Garimidi, Yang Liu
Accurate prediction of the critical heat flux (CHF) is a crucial design and safety consideration for a wide range of high-performance thermal systems, including water-cooled nuclear reactors. Traditional predictive tools, such as empirical correlations and look-up tables, often lack accuracy when extrapolated or at different interpolation regions. To overcome these limitations, this work introduces a novel physics-guided, optimized deep-ensemble framework for robust CHF prediction with comprehensive uncertainty quantification. Our approach first expands the model’s inputs by augmenting base thermal-hydraulic parameters with physics-based features derived from established correlations. This feature engineering injects domain knowledge, constraining the solution space and promoting convergence to physically plausible solutions. Furthermore, we employ a sophisticated hyperparameter optimization strategy, combining a Sobol sequence with Bayesian optimization, to systematically select a diverse and high-performing set of neural networks for the ensemble. The resulting physics-guided ensemble demonstrates superior performance across all metrics compared to a baseline ensemble, a standard look-up table, and a benchmark neural network. The model produces smoother, more physically consistent predictive trends and provides reliable uncertainty estimates. This framework offers a powerful and broadly applicable tool for CHF prediction, enabling higher-fidelity safety margins and the design of more efficient and reliable thermal management systems.
{"title":"Bayesian-optimized, feature-augmented deep ensemble for physics-guided critical heat-flux prediction with uncertainty quantification","authors":"Zaid Abulawi, Doyeong Lim, Abhiram Garimidi, Yang Liu","doi":"10.1016/j.anucene.2026.112139","DOIUrl":"10.1016/j.anucene.2026.112139","url":null,"abstract":"<div><div>Accurate prediction of the critical heat flux (CHF) is a crucial design and safety consideration for a wide range of high-performance thermal systems, including water-cooled nuclear reactors. Traditional predictive tools, such as empirical correlations and look-up tables, often lack accuracy when extrapolated or at different interpolation regions. To overcome these limitations, this work introduces a novel physics-guided, optimized deep-ensemble framework for robust CHF prediction with comprehensive uncertainty quantification. Our approach first expands the model’s inputs by augmenting base thermal-hydraulic parameters with physics-based features derived from established correlations. This feature engineering injects domain knowledge, constraining the solution space and promoting convergence to physically plausible solutions. Furthermore, we employ a sophisticated hyperparameter optimization strategy, combining a Sobol sequence with Bayesian optimization, to systematically select a diverse and high-performing set of neural networks for the ensemble. The resulting physics-guided ensemble demonstrates superior performance across all metrics compared to a baseline ensemble, a standard look-up table, and a benchmark neural network. The model produces smoother, more physically consistent predictive trends and provides reliable uncertainty estimates. This framework offers a powerful and broadly applicable tool for CHF prediction, enabling higher-fidelity safety margins and the design of more efficient and reliable thermal management systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112139"},"PeriodicalIF":2.3,"publicationDate":"2026-01-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145975286","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 characteristics of a born-free core in an integrated natural circulation SMR have been researched in the present study. The coupled model of SMR is established by integrating nuclear design code CMS-5, thermal–hydraulic analysis code RELAP5 and subchannel code. Integrated multi-physics analysis reveals intrinsic feedback mechanisms governing flow-power synchronization in the closed parallel channels. It demonstrates that natural circulation flow distribution exhibits self-adaptive proportionality to assemblies power distribution. Besides control of the power distribution by nuclear design, adjusting structural parameters, such as increasing the closed riser height or optimizing the inlet resistance coefficient, can improve flow distribution, reduce outlet temperature differences and enhance thermal performance. Furthermore, minor subcooled boiling occurs in higher-power assemblies, producing bubbles that increase the driving force of natural circulation. However, if the power factor is too high, the large number of bubbles may cause overall natural circulation flow to fluctuate, despite sufficient subcooling after coolant mixing.
{"title":"Characteristics of a boron-free zirconium boxed core in an integrated natural circulation SMR","authors":"Yuhong Wang, Ting Wei, Zhidong Yue, Zhiyong Li, Ying Zhang","doi":"10.1016/j.anucene.2026.112121","DOIUrl":"10.1016/j.anucene.2026.112121","url":null,"abstract":"<div><div>The characteristics of a born-free core in an integrated natural circulation SMR have been researched in the present study. The coupled model of SMR is established by integrating nuclear design code CMS-5, thermal–hydraulic analysis code RELAP5 and subchannel<!--> <!-->code. Integrated multi-physics analysis reveals intrinsic feedback mechanisms governing flow-power synchronization in the closed parallel channels. It demonstrates that natural circulation flow distribution exhibits self-adaptive proportionality to assemblies power distribution. Besides control of the power distribution by nuclear design, adjusting structural parameters, such as increasing the closed riser height or optimizing the inlet resistance coefficient, can improve flow distribution, reduce outlet temperature differences and enhance thermal performance. Furthermore, minor subcooled boiling occurs in higher-power assemblies, producing bubbles that increase the driving force of natural circulation. However, if the power factor is too high, the large number of bubbles may cause overall natural circulation flow to fluctuate, despite sufficient subcooling after coolant mixing.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112121"},"PeriodicalIF":2.3,"publicationDate":"2026-01-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145975287","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}
Fracture behavior of chromium (Cr) coated claddings under loss of coolant accident (LOCA) conditions were investigated utilizing the FEMAXI fuel performance code with newly implemented Cr coating degradation models. The FEMAXI code reproduced microstructure evolution and cladding oxidation under LOCA conditions, including metallic and ZrO2 layers growth and oxygen uptake. Sensitivity analyses of the cladding oxygen concentration, where the effects of wall thickness change and eutectic reactions were taken into account, indicate that the fracture condition of the Cr-coated cladding can be discriminated by a criterion based on the remaining β-Zr thickness with an oxygen concentration of ≤ 0.9 wt%. This demonstrates FEMAXI’s applicability for assessing Cr-coated cladding performance under accident scenarios.
{"title":"Analysis of fracture conditions of Cr-coated Zr alloy claddings under LOCA conditions calculated using FEMAXI fuel performance code","authors":"Vu-Nhut Luu, Yoshinori Taniguchi, Yutaka Udagawa, Yudai Tasaki, Jinya Katsuyama","doi":"10.1016/j.anucene.2026.112114","DOIUrl":"10.1016/j.anucene.2026.112114","url":null,"abstract":"<div><div>Fracture behavior of chromium (Cr) coated claddings under loss of coolant accident (LOCA) conditions were investigated utilizing the FEMAXI fuel performance code with newly implemented Cr coating degradation models. The FEMAXI code reproduced microstructure evolution and cladding oxidation under LOCA conditions, including metallic and ZrO<sub>2</sub> layers growth and oxygen uptake. Sensitivity analyses of the cladding oxygen concentration, where the effects of wall thickness change and eutectic reactions were taken into account, indicate that the fracture condition of the Cr-coated cladding can be discriminated by a criterion based on the remaining β-Zr thickness with an oxygen concentration of ≤ 0.9 wt%. This demonstrates FEMAXI’s applicability for assessing Cr-coated cladding performance under accident scenarios.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112114"},"PeriodicalIF":2.3,"publicationDate":"2026-01-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145975066","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 : 2026-01-10DOI: 10.1016/j.anucene.2026.112131
Achu Govind K.R.
Reliable reactive power regulation in Molten Salt Breeder Reactor (MSBR) cores is essential for safety and efficiency. Conventional controllers often exhibit poor robustness and fault tolerance under uncertainties, nonlinear dynamics, and disturbances. To address these limitations, this work introduces a Quantitative Feedback Theory (QFT)-based proportional integral derivative (PID) controller enhanced with Long Short-Term Memory (LSTM) modeling. The controller is designed in two stages. First, a QFT-based PID controller is synthesized to guarantee stability and robust performance across plant uncertainty sets. Robustness, disturbance rejection, and tracking are enforced as frequency-domain inequalities. In the second stage, an LSTM network is integrated to adaptively tune PID gains in real time. This ensures that predictions inherently satisfy robust stability, tracking, and disturbance rejection constraints. A composite performance-driven loss further biases the network toward minimizing integral absolute error (IAE), overshoot, and settling time while preserving robustness. The QFT-PID-LSTM controller achieved faster rise time, shorter settling time, negligible steady-state error, and lower control effort compared to existing approaches. Quantitative indices showed reductions of nearly 90% in performance metrics and 80% in statistical measures. Disk margin analysis confirmed stability, while Monte Carlo simulations demonstrated tightly bounded error distributions. The controller also maintained stability and accurate tracking under both sensor and actuator faults, confirming strong robustness and fault tolerance.
{"title":"Robust and fault-tolerant control of MSBR reactor using a hybrid QFT–PID–LSTM framework with disk margin analysis","authors":"Achu Govind K.R.","doi":"10.1016/j.anucene.2026.112131","DOIUrl":"10.1016/j.anucene.2026.112131","url":null,"abstract":"<div><div>Reliable reactive power regulation in Molten Salt Breeder Reactor (MSBR) cores is essential for safety and efficiency. Conventional controllers often exhibit poor robustness and fault tolerance under uncertainties, nonlinear dynamics, and disturbances. To address these limitations, this work introduces a Quantitative Feedback Theory (QFT)-based proportional integral derivative (PID) controller enhanced with Long Short-Term Memory (LSTM) modeling. The controller is designed in two stages. First, a QFT-based PID controller is synthesized to guarantee stability and robust performance across plant uncertainty sets. Robustness, disturbance rejection, and tracking are enforced as frequency-domain inequalities. In the second stage, an LSTM network is integrated to adaptively tune PID gains in real time. This ensures that predictions inherently satisfy robust stability, tracking, and disturbance rejection constraints. A composite performance-driven loss further biases the network toward minimizing integral absolute error (IAE), overshoot, and settling time while preserving robustness. The QFT-PID-LSTM controller achieved faster rise time, shorter settling time, negligible steady-state error, and lower control effort compared to existing approaches. Quantitative indices showed reductions of nearly 90% in performance metrics and 80% in statistical measures. Disk margin analysis confirmed stability, while Monte Carlo simulations demonstrated tightly bounded error distributions. The controller also maintained stability and accurate tracking under both sensor and actuator faults, confirming strong robustness and fault tolerance.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112131"},"PeriodicalIF":2.3,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145924528","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 : 2026-01-10DOI: 10.1016/j.anucene.2025.112105
Mohamed Hasabelnaby , Mohammad Marashdeh , K.A. Mahmoud , Reham M. Abd El Rahman , Hanan Akhdar , Ghada Salaheldin , Mohammad Y. Hanfi
Radiation shielding materials are crucial for nuclear, industrial, and medical situations where shielding against ionizing radiation is a concern. In this study, the effect of ferric oxide (Fe2O3) incorporation on the mineralogical, physical, mechanical, and gamma-ray shielding properties of Portland cement-based concretes is studied. Concrete samples were made by replacing fine aggregate with Fe2O3 (0–40 wt%) in various amounts and were examined for mineralogy using XRD, elemental composition using XRF, density, porosity, and water absorption, compressive strength, elastic modulus, and compression through gamma-ray attenuation. The results showed that on average Fe2O3 incorporation led to higher concrete density (2.51–2.69 g/cm3), lower porosity of concrete, and a more than 38 % and 36 % reduction in water absorption, and improved gamma-shielding performance, with average increases in linear attenuation coefficient (LAC) of ∼7 % at energies of 0.511 and 0.662 MeV. Nevertheless, average compressive strength declined from 9.75 MPa (control) to 3.75 MPa (40 wt% Fe2O3) and the elastic modulus from 15.6 GPa to 9.7 GPa which are not strong load bearing results. Regression analysis produced predictive models (R2 > 0.95) relating Fe2O3 amount to density, porosity, and strength to allow for performance estimation for design. These results confirmed that Fe2O3 concretes, while not viable for structural load bearing, would still serve as effective non-structural shielding materials for medical and nuclear applications. Based on the data obtained from the Fe2O3 study, the upper limit of enhancement through Fe2O3 was assigned to CON30 to CON40, though these levels represent some of the highest attenuation values detected in relation to mechanical degradation. Therefore, the Fe2O3 doping group concretes do not uphold their suitability for load bearing applications but do give considerable merit for use as non-load bearing radiation shielding materials across medical, research, and nuclear facilities.
{"title":"Influence of ferric oxide (Fe2O3) content on the mechanical strength and radiation attenuation capacity of concrete","authors":"Mohamed Hasabelnaby , Mohammad Marashdeh , K.A. Mahmoud , Reham M. Abd El Rahman , Hanan Akhdar , Ghada Salaheldin , Mohammad Y. Hanfi","doi":"10.1016/j.anucene.2025.112105","DOIUrl":"10.1016/j.anucene.2025.112105","url":null,"abstract":"<div><div>Radiation shielding materials are crucial for nuclear, industrial, and medical situations where shielding against ionizing radiation is a concern. In this study, the effect of ferric oxide (Fe<sub>2</sub>O<sub>3</sub>) incorporation on the mineralogical, physical, mechanical, and gamma-ray shielding properties of Portland cement-based concretes is studied. Concrete samples were made by replacing fine aggregate with Fe<sub>2</sub>O<sub>3</sub> (0–40 wt%) in various amounts and were examined for mineralogy using XRD, elemental composition using XRF, density, porosity, and water absorption, compressive strength, elastic modulus, and compression through gamma-ray attenuation. The results showed that on average Fe<sub>2</sub>O<sub>3</sub> incorporation led to higher concrete density (2.51–2.69 g/cm<sup>3</sup>), lower porosity of concrete, and a more than 38 % and 36 % reduction in water absorption, and improved gamma-shielding performance, with average increases in linear attenuation coefficient (LAC) of ∼7 % at energies of 0.511 and 0.662 MeV. Nevertheless, average compressive strength declined from 9.75 MPa (control) to 3.75 MPa (40 wt% Fe<sub>2</sub>O<sub>3</sub>) and the elastic modulus from 15.6 GPa to 9.7 GPa which are not strong load bearing results. Regression analysis produced predictive models (R<sup>2</sup> > 0.95) relating Fe<sub>2</sub>O<sub>3</sub> amount to density, porosity, and strength to allow for performance estimation for design. These results confirmed that Fe<sub>2</sub>O<sub>3</sub> concretes, while not viable for structural load bearing, would still serve as effective non-structural shielding materials for medical and nuclear applications. Based on the data obtained from the Fe<sub>2</sub>O<sub>3</sub> study, the upper limit of enhancement through Fe<sub>2</sub>O<sub>3</sub> was assigned to CON30 to CON40, though these levels represent some of the highest attenuation values detected in relation to mechanical degradation. Therefore, the Fe<sub>2</sub>O<sub>3</sub> doping group concretes do not uphold their suitability for load bearing applications but do give considerable merit for use as non-load bearing radiation shielding materials across medical, research, and nuclear facilities.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112105"},"PeriodicalIF":2.3,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145924527","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 : 2026-01-09DOI: 10.1016/j.anucene.2026.112122
Jingxiang Cao , Guifeng Zhu , Huiqin Yin , Linbing Jiang , Xinmei Yang , Jie Qiu , Wenguan Liu
Tellurium (Te) is the primary cause of intergranular embrittlement in structural materials of molten salt reactors (MSR). This study investigates the chemical behavior of Te in FLiBe molten salt under different redox states, and finds that the redox states have a substantial impact on the chemical behavior of Te. Under reducing redox conditions of the molten salt, Te can stably exist as an anion and preferentially forms bonds with positively charged Th or U atoms. In neutral or mildly oxidizing environments, Te atoms are more likely to aggregate and form Te–Te bonds, which facilitates nucleation in the molten salt and promotes their adsorption onto the alloy surface. Under strongly oxidizing conditions, Te tends to exist in a cationic form and may be present as tellurium fluoride gas. This study reveals the possibility of inhibiting Te-induced intergranular embrittlement in MSRs by adjusting the redox state of the molten salt.
{"title":"Effect of molten salt redox states on the chemical behavior of Tellurium: A machine learning molecular dynamics study","authors":"Jingxiang Cao , Guifeng Zhu , Huiqin Yin , Linbing Jiang , Xinmei Yang , Jie Qiu , Wenguan Liu","doi":"10.1016/j.anucene.2026.112122","DOIUrl":"10.1016/j.anucene.2026.112122","url":null,"abstract":"<div><div>Tellurium (Te) is the primary cause of intergranular embrittlement in structural materials of molten salt reactors (MSR). This study investigates the chemical behavior of Te in FLiBe molten salt under different redox states, and finds that the redox states have a substantial impact on the chemical behavior of Te. Under reducing redox conditions of the molten salt, Te can stably exist as an anion and preferentially forms bonds with positively charged Th or U atoms. In neutral or mildly oxidizing environments, Te atoms are more likely to aggregate and form Te–Te bonds, which facilitates nucleation in the molten salt and promotes their adsorption onto the alloy surface. Under strongly oxidizing conditions, Te tends to exist in a cationic form and may be present as tellurium fluoride gas. This study reveals the possibility of inhibiting Te-induced intergranular embrittlement in MSRs by adjusting the redox state of the molten salt.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112122"},"PeriodicalIF":2.3,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145915165","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 : 2026-01-09DOI: 10.1016/j.anucene.2025.112111
Shikhar Kumar , Changho Lee , Vincent Laboure , Yeon Sang Jung , Stefano Terlizzi , Yaqi Wang , Javier Ortensi
This work introduces the pin power reconstruction capability available in the Griffin reactor physics code. This capability is implemented in an unstructured mesh framework, and the methods introduced are applied to the 2D SIMBA reactor core, which has assemblies and pins arranged in a hexagonal lattice. Since this reactor has a non-Cartesian geometry and also operates in the thermal spectrum, a general approach to pin power reconstruction is adopted, where SPH-based equivalence is leveraged to preserve assembly-wise reaction rates, while computing full-core form functions to preserve pin-wise fission production rates within the fuel pins of the reactor core. In a 2D microreactor benchmark problem, this pin power reconstruction approach was shown to reproduce pin powers compared to the Serpent2 Monte Carlo code for fixed temperature conditions and control drum rotation angles, yielding a core-wide RMS error level of 0.6% and a maximum absolute pin error of 2.3%. In addition, a tabulated library of multigroup cross sections, SPH factors, and form functions was generated to demonstrate the applicability of pin power reconstruction to a thermal feedback problem. Finally, a control drum transient was successfully simulated, showcasing the application of pin power reconstruction in a transient multiphysics feedback problem.
{"title":"Transient multiphysics simulations with pin power reconstruction in the Griffin reactor physics code","authors":"Shikhar Kumar , Changho Lee , Vincent Laboure , Yeon Sang Jung , Stefano Terlizzi , Yaqi Wang , Javier Ortensi","doi":"10.1016/j.anucene.2025.112111","DOIUrl":"10.1016/j.anucene.2025.112111","url":null,"abstract":"<div><div>This work introduces the pin power reconstruction capability available in the Griffin reactor physics code. This capability is implemented in an unstructured mesh framework, and the methods introduced are applied to the 2D SIMBA reactor core, which has assemblies and pins arranged in a hexagonal lattice. Since this reactor has a non-Cartesian geometry and also operates in the thermal spectrum, a general approach to pin power reconstruction is adopted, where SPH-based equivalence is leveraged to preserve assembly-wise reaction rates, while computing full-core form functions to preserve pin-wise fission production rates within the fuel pins of the reactor core. In a 2D microreactor benchmark problem, this pin power reconstruction approach was shown to reproduce pin powers compared to the Serpent2 Monte Carlo code for fixed temperature conditions and control drum rotation angles, yielding a core-wide RMS error level of 0.6% and a maximum absolute pin error of 2.3%. In addition, a tabulated library of multigroup cross sections, SPH factors, and form functions was generated to demonstrate the applicability of pin power reconstruction to a thermal feedback problem. Finally, a control drum transient was successfully simulated, showcasing the application of pin power reconstruction in a transient multiphysics feedback problem.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112111"},"PeriodicalIF":2.3,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145915172","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 : 2026-01-09DOI: 10.1016/j.anucene.2026.112120
Jiaqing Zhang, Xiao Zhang, Wenxiao Chu, Qiuwang Wang
Heat pipe cooled reactors (HPCRs) rely on thermal conduction to transfer heat from the reactor core, where the thermal reliability becomes a critical concern. Studies on the temperature dynamic response due to random heat pipe (HP) failures and the prediction of specific heat pipes through temperature distribution analysis are the key challenges. This study investigates the spatial thermal diffusion mechanism and temperature dynamic response characteristics in the HPCR core during random HP failure processes using experimental and numerical methods. Moreover, the Random Forest algorithm method is introduced to predict HP failure locations. Results indicate that boundary HP failure (HP-A) exhibits a broader critical failure diffusion radius of 65.1 mm and diffusion angle of 190°, while central HP failure (HP-D) causes minimal disturbance and results in more uniform temperature gradient distributions. Correspondingly, the dynamic response time constant and response delay time are employed to quantitatively feature the temperature field evolution during HP failure. For HP-A, the time constant and response delay time are 5040 s and 170 s, respectively, compared to 10,950 and 550 s for HP-D. Additionally, two patterns with single and dual HP failures and four HP failure orientations are predicted by the Random Forest algorithm method. Results demonstrate the prediction accuracy of 97.1 %, with a failure time prediction error ranging from −0.7 % to 1.6 %.
{"title":"Transient thermal diffusion analysis and failure prediction in heat-pipe-cooled reactors","authors":"Jiaqing Zhang, Xiao Zhang, Wenxiao Chu, Qiuwang Wang","doi":"10.1016/j.anucene.2026.112120","DOIUrl":"10.1016/j.anucene.2026.112120","url":null,"abstract":"<div><div>Heat pipe cooled reactors (HPCRs) rely on thermal conduction to transfer heat from the reactor core, where the thermal reliability becomes a critical concern. Studies on the temperature dynamic response due to random heat pipe (HP) failures and the prediction of specific heat pipes through temperature distribution analysis are the key challenges. This study investigates the spatial thermal diffusion mechanism and temperature dynamic response characteristics in the HPCR core during random HP failure processes using experimental and numerical methods. Moreover, the Random Forest algorithm method is introduced to predict HP failure locations. Results indicate that boundary HP failure (HP-A) exhibits a broader critical failure diffusion radius of 65.1 mm and diffusion angle of 190°, while central HP failure (HP-D) causes minimal disturbance and results in more uniform temperature gradient distributions. Correspondingly, the dynamic response time constant and response delay time are employed to quantitatively feature the temperature field evolution during HP failure. For HP-A, the time constant and response delay time are 5040 s and 170 s, respectively, compared to 10,950 and 550 s for HP-D. Additionally, two patterns with single and dual HP failures and four HP failure orientations are predicted by the Random Forest algorithm method. Results demonstrate the prediction accuracy of 97.1 %, with a failure time prediction error ranging from −0.7 % to 1.6 %.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112120"},"PeriodicalIF":2.3,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145915169","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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