Pub Date : 2026-06-01Epub Date: 2026-01-24DOI: 10.1016/j.anucene.2026.112168
Yifan Zhou, Houjian Zhao, Yang Liu
Shell and tube heat exchangers are widely used in nuclear engineering and the petrochemical industries. In the current investigation, cross flow over inline tube bundles with various pitch-to-diameter ratios is simulated by SST k-ω-γ. The mesh near the shear layer region is refined due to the large velocity gradient. The effects of the bounding wall, end wall, and pitch ratio on time-averaged and transient flow fields are systematically analyzed. The increased streamwise pitch results in the impinging point shifting to near θ = 0°. The increased transverse pitch results in a larger influence on side passages. The recirculation region near the end wall is attenuated, resulting in reduced drag and large velocity magnitude. The separated vortices after the middle tubes sway into the main flow. There is a separation vortex near the bounding wall due to the entrainment of separate vortices.
{"title":"Numerical investigation of wall effects on cross flow over inline tube bundles with various pitch-to-diameter ratios","authors":"Yifan Zhou, Houjian Zhao, Yang Liu","doi":"10.1016/j.anucene.2026.112168","DOIUrl":"10.1016/j.anucene.2026.112168","url":null,"abstract":"<div><div>Shell and tube heat exchangers are widely used in nuclear engineering and the petrochemical industries. In the current investigation, cross flow over inline tube bundles with various pitch-to-diameter ratios is simulated by SST <em>k-ω-γ.</em> The mesh near the shear layer region is refined due to the large velocity gradient. The effects of the bounding wall, end wall, and pitch ratio on time-averaged and transient flow fields are systematically analyzed. The increased streamwise pitch results in the impinging point shifting to near <em>θ</em> = 0°. The increased transverse pitch results in a larger influence on side passages. The recirculation region near the end wall is attenuated, resulting in reduced drag and large velocity magnitude. The separated vortices after the middle tubes sway into the main flow. There is a separation vortex near the bounding wall due to the entrainment of separate vortices.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112168"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035468","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 : 2026-06-01Epub Date: 2026-01-22DOI: 10.1016/j.anucene.2026.112117
Peng Sitao, Wei Jinfeng, Yao Jianfan, Yang Shuoyan, Li Wenhuai, Yang Zhengyu, Huang Jie, Mei Gaohui, Huang Yunten, Zhao Changyoug, Yu Chao, Wang Ting, Li Jinggang
Commercial pressurized water reactor (PWR) nuclear power plants require loading pattern (LP) design every 12–24 months. To address the time-consuming nature of routine loading pattern searches and their strong reliance on engineer expertise, this study proposes an AI-based automated loading pattern search method and develops the corresponding intelligent loading pattern design software KAPOK. This software automates tasks previously requiring experienced engineers through automation and intelligent technologies. KAPOK employs semi-empirical methods for automatic selection of spent fuel assemblies, heuristic algorithms for pattern search, and integrates neural network surrogate models for rapid assessment of key refueling parameters with the comprehensive evaluation capabilities of nuclear design software. Typically, KAPOK can identify highly optimized solutions within 30 min, achieving exceptional efficiency through automated processing. Validated extensively against historical cycle patterns, KAPOK has been successfully deployed in routine refueling designs at nuclear power plants.
{"title":"Research and application of intelligent loading pattern search for pressurized water reactors","authors":"Peng Sitao, Wei Jinfeng, Yao Jianfan, Yang Shuoyan, Li Wenhuai, Yang Zhengyu, Huang Jie, Mei Gaohui, Huang Yunten, Zhao Changyoug, Yu Chao, Wang Ting, Li Jinggang","doi":"10.1016/j.anucene.2026.112117","DOIUrl":"10.1016/j.anucene.2026.112117","url":null,"abstract":"<div><div>Commercial pressurized water reactor (PWR) nuclear power plants require loading pattern (LP) design every 12–24 months. To address the time-consuming nature of routine loading pattern searches and their strong reliance on engineer expertise, this study proposes an AI-based automated loading pattern search method and develops the corresponding intelligent loading pattern design software KAPOK. This software automates tasks previously requiring experienced engineers through automation and intelligent technologies. KAPOK employs semi-empirical methods for automatic selection of spent fuel assemblies, heuristic algorithms for pattern search, and integrates neural network surrogate models for rapid assessment of key refueling parameters with the comprehensive evaluation capabilities of nuclear design software. Typically, KAPOK can identify highly optimized solutions within 30 min, achieving exceptional efficiency through automated processing. Validated extensively against historical cycle patterns, KAPOK has been successfully deployed in routine refueling designs at nuclear power plants.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112117"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035539","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 : 2026-06-01Epub Date: 2026-01-22DOI: 10.1016/j.anucene.2026.112130
Qiming Yang , Yuan Yuan , Dong Huang , Youqi Zheng , Guoming Liu
This paper presents SaraGR-K, a three-dimensional code developed for coupled neutronics/thermal-hydraulics (N/T-H) transient analysis, specifically designed for prismatic gas-cooled microreactors. The code solves the 3D neutron transport equation in the time–space domain using a predictor–corrector quasi-static (PCQS) method. To further enhance computational efficiency in 3D transport simulations, SaraGR-K incorporates a fixed-source scaling factor technique to accelerate the convergence of transient fixed-source equations and utilizes a pipelined parallelization strategy to expedite transport calculations. Temperature distributions within reactor materials are subsequently calculated using a 1D fluid dynamics model coupled with a 1D solid heat conduction model, ensuring consistent integration with the neutronics solution.
With these computational and thermal–hydraulic models in place, SaraGR-K was verified using numerical benchmarks problems and by comparing its results with those from established reference codes. Following the verification, it was applied to the transient analysis of a prismatic gas-cooled microreactor, demonstrating its feasibility for full-core transient simulations in such systems.
{"title":"SaraGR-K: a deterministic transport-based transient analysis code for prismatic gas-cooled micro-reactors","authors":"Qiming Yang , Yuan Yuan , Dong Huang , Youqi Zheng , Guoming Liu","doi":"10.1016/j.anucene.2026.112130","DOIUrl":"10.1016/j.anucene.2026.112130","url":null,"abstract":"<div><div>This paper presents SaraGR-K, a three-dimensional code developed for coupled neutronics/thermal-hydraulics (N/T-H) transient analysis, specifically designed for prismatic gas-cooled microreactors. The code solves the 3D neutron transport equation in the time–space domain using a predictor–corrector quasi-static (PCQS) method. To further enhance computational efficiency in 3D transport simulations, SaraGR-K incorporates a fixed-source scaling factor technique to accelerate the convergence of transient fixed-source equations and utilizes a pipelined parallelization strategy to expedite transport calculations. Temperature distributions within reactor materials are subsequently calculated using a 1D fluid dynamics model coupled with a 1D solid heat conduction model, ensuring consistent integration with the neutronics solution.</div><div>With these computational and thermal–hydraulic models in place, SaraGR-K was verified using numerical benchmarks problems and by comparing its results with those from established reference codes. Following the verification, it was applied to the transient analysis of a prismatic gas-cooled microreactor, demonstrating its feasibility for full-core transient simulations in such systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112130"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035537","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 : 2026-06-01Epub Date: 2026-01-21DOI: 10.1016/j.anucene.2026.112132
Restu Kojo
In risk-informed regulation, the significance of systems, structures, and components (SSCs) is assessed using multiple risk metrics, requiring a systematic method to determine whether SSC degradation has a greater impact on Level 1 or Level 2 probabilistic risk assessments (PRA). A key issue is that risk significance in Level 2 often exceeds that of Level 1 PRA due to the order-of-magnitude difference between target values for core damage frequency (CDF) and that of containment failure frequency (CFF). To address this, a new methodology was developed, including a novel measure—risk difference achievement worth (RDAW)—which enables transparent comparisons across different PRAs. This methodology was applied to large-scale PRA models and confirmed the consistency of the significance comparison results. In summary, a mathematically formulated methodology for comparing significance across multiple PRAs, which is applicable to large-scale practical models, has been established.
{"title":"Methodology for significance determination across multiple risk metrics using novel importance measures","authors":"Restu Kojo","doi":"10.1016/j.anucene.2026.112132","DOIUrl":"10.1016/j.anucene.2026.112132","url":null,"abstract":"<div><div>In risk-informed regulation, the significance of systems, structures, and components (SSCs) is assessed using multiple risk metrics, requiring a systematic method to determine whether SSC degradation has a greater impact on Level 1 or Level 2 probabilistic risk assessments (PRA). A key issue is that risk significance in Level 2 often exceeds that of Level 1 PRA due to the order-of-magnitude difference between target values for core damage frequency (CDF) and that of containment failure frequency (CFF). To address this, a new methodology was developed, including a novel measure—risk difference achievement worth (RDAW)—which enables transparent comparisons across different PRAs. This methodology was applied to large-scale PRA models and confirmed the consistency of the significance comparison results. In summary, a mathematically formulated methodology for comparing significance across multiple PRAs, which is applicable to large-scale practical models, has been established.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112132"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035531","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 : 2026-06-01Epub Date: 2026-02-12DOI: 10.1016/j.anucene.2026.112204
Abdullah Zengin , Ekrem Gulsevincler
The operational readiness and construction process of Akkuyu NGS, Türkiye’s first nuclear power plant, is ongoing. This situation necessitates atmospheric accident analyses for the nuclear power plant. While such accident analyses are generally available for nuclear facilities, the fact that one side of the Akkuyu site is surrounded by sea and the other by mountains makes it critical to perform radiological dispersion analyses that take into account different atmospheric stability conditions. The change and analysis of the total effective dose equivalent (TEDE), respirable time-integrated air concentration (RTIAC), and ground deposition parameters of the pollutant released after a plant accident under different stability classes is of critical importance in emergency planning. Using local meteorological data from 2019 to 2023 and the HotSpot dispersion code, key radiological parameters such as Total Effective Dose Equivalent (TEDE), respirable air concentration (RTIAC), and ground deposition were evaluated. The findings reveal a critical situation determined by atmospheric stability. In the highly unstable Class A, it causes severe and high dose exposure near the plant area, while in Class D analyses, dose exposure is at lower concentrations, but the spread affects a wider geographical area. These results highlight the importance of a specially designed emergency strategy for the Akkuyu NGS site that distinguishes between Class A emergency local response and Class D large-scale intervention and protective actions.
{"title":"Radiological assessment of a potential accident scenario at the Akkuyu Nuclear Power Plant: TEDE, respirable time integrated air concentration, and ground surface deposition under different stability classes","authors":"Abdullah Zengin , Ekrem Gulsevincler","doi":"10.1016/j.anucene.2026.112204","DOIUrl":"10.1016/j.anucene.2026.112204","url":null,"abstract":"<div><div>The operational readiness and construction process of Akkuyu NGS, Türkiye’s first nuclear power plant, is ongoing. This situation necessitates atmospheric accident analyses for the nuclear power plant. While such accident analyses are generally available for nuclear facilities, the fact that one side of the Akkuyu site is surrounded by sea and the other by mountains makes it critical to perform radiological dispersion analyses that take into account different atmospheric stability conditions. The change and analysis of the total effective dose equivalent (TEDE), respirable time-integrated air concentration (RTIAC), and ground deposition parameters of the pollutant released after a plant accident under different stability classes is of critical importance in emergency planning. Using local meteorological data from 2019 to 2023 and the HotSpot dispersion code, key radiological parameters such as Total Effective Dose Equivalent (TEDE), respirable air concentration (RTIAC), and ground deposition were evaluated. The findings reveal a critical situation determined by atmospheric stability. In the highly unstable Class A, it causes severe and high dose exposure near the plant area, while in Class D analyses, dose exposure is at lower concentrations, but the spread affects a wider geographical area. These results highlight the importance of a specially designed emergency strategy for the Akkuyu NGS site that distinguishes between Class A emergency local response and Class D large-scale intervention and protective actions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112204"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186623","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 : 2026-06-01Epub Date: 2026-02-09DOI: 10.1016/j.anucene.2026.112196
Yusen Liu , Sinan Chen , Yifan Li , Xiangyi Du , Zhiyi Wang , Xiumei Wang , Hanbao Chong , Mingzhang Lin
The synthesis and utilization of sulfonic acid-functionalized graphitic carbon nitride (C3N4-SO3H) as an effective adsorbent for the removal of Co2+ from aqueous solutions was successful in varying experimental conditions, such as solution pH, adsorption time, and temperature. Kinetic studies revealed that the adsorption process is largely determined by chemical interactions. The experimental data was fitted to Langmuir and Freundlich adsorption isotherm models, and the results showed that Co2+ adsorption onto C3N4-SO3H is a monolayer process. The maximum adsorption capacity of C3N4-SO3H for Co2+ was found to be 247.52 mg g−1, which is significantly higher than that of unmodified graphitic carbon nitride (g-C3N4). An analysis of thermodynamics proposed that the adsorption process is both spontaneous and endothermic. Furthermore, the adsorption mechanism of Co2+ by C3N4-SO3H was investigated using density functional theory (DFT) calculations, which confirmed that the adsorption process is energetically favorable and spontaneous. The adsorption of Co2+ by C3N4-SO3H was demonstrated to be strong, even under irradiation and high temperatures, thus highlighting its potential for use in the elimination of Co2+ from radioactive wastewater.
{"title":"The effective adsorption of cobalt from radioactive wastewater by sulfonated g-C3N4","authors":"Yusen Liu , Sinan Chen , Yifan Li , Xiangyi Du , Zhiyi Wang , Xiumei Wang , Hanbao Chong , Mingzhang Lin","doi":"10.1016/j.anucene.2026.112196","DOIUrl":"10.1016/j.anucene.2026.112196","url":null,"abstract":"<div><div>The synthesis and utilization of sulfonic acid-functionalized graphitic carbon nitride (C<sub>3</sub>N<sub>4</sub>-SO<sub>3</sub>H) as an effective adsorbent for the removal of Co<sup>2+</sup> from aqueous solutions was successful in varying experimental conditions, such as solution pH, adsorption time, and temperature. Kinetic studies revealed that the adsorption process is largely determined by chemical interactions. The experimental data was fitted to Langmuir and Freundlich adsorption isotherm models, and the results showed that Co<sup>2+</sup> adsorption onto C<sub>3</sub>N<sub>4</sub>-SO<sub>3</sub>H is a monolayer process. The maximum adsorption capacity of C<sub>3</sub>N<sub>4</sub>-SO<sub>3</sub>H for Co<sup>2+</sup> was found to be 247.52 mg g<sup>−1</sup>, which is significantly higher than that of unmodified graphitic carbon nitride (g-C<sub>3</sub>N<sub>4</sub>). An analysis of thermodynamics proposed that the adsorption process is both spontaneous and endothermic. Furthermore, the adsorption mechanism of Co<sup>2+</sup> by C<sub>3</sub>N<sub>4</sub>-SO<sub>3</sub>H was investigated using density functional theory (DFT) calculations, which confirmed that the adsorption process is energetically favorable and spontaneous. The adsorption of Co<sup>2+</sup> by C<sub>3</sub>N<sub>4</sub>-SO<sub>3</sub>H was demonstrated to be strong, even under irradiation and high temperatures, thus highlighting its potential for use in the elimination of Co<sup>2+</sup> from radioactive wastewater.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112196"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186638","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 : 2026-06-01Epub 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":"2026-06-01","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}
As energy structures evolve and environmental standards rise, nuclear energy shows clear advantages in district heating, particularly in cold regions. To ensure consistent quality and stability of supply, accurate and responsive load-following capability is essential for the effective use of nuclear energy in the heating sector.
This study presents a simulation model for a 400 MW pool-type nuclear heating reactor, accompanied by the development of a comprehensive system simulation platform using Python and Computational Fluid Dynamics (CFD) for both one-dimensional (1D) and three-dimensional (3D) coupled analyses. The research systematically examines the operational parameters and strategies of the nuclear heating system under extreme cold climate conditions. The simulation results indicate that under the constant-temperature heating mode, after accounting for the thermal delay characteristics of the reactor pool, the number of power adjustments required during the 139-day heating season is reduced by 22 instances. Under the variable-temperature heating mode, the independent heating configuration of this reactor can satisfy the thermal demand during the heating period in cold regions.
{"title":"Operational strategies for nuclear district heating systems in extremely cold climates","authors":"Ruo-Jun Xue , Han-Wen Liang , Hao-Fang Chong , Min-Jun Peng","doi":"10.1016/j.anucene.2026.112162","DOIUrl":"10.1016/j.anucene.2026.112162","url":null,"abstract":"<div><div>As energy structures evolve and environmental standards rise, nuclear energy shows clear advantages in district heating, particularly in cold regions. To ensure consistent quality and stability of supply, accurate and responsive load-following capability is essential for the effective use of nuclear energy in the heating sector.</div><div>This study presents a simulation model for a 400 MW pool-type nuclear heating reactor, accompanied by the development of a comprehensive system simulation platform using Python and Computational Fluid Dynamics (CFD) for both one-dimensional (1D) and three-dimensional (3D) coupled analyses. The research systematically examines the operational parameters and strategies of the nuclear heating system under extreme cold climate conditions. The simulation results indicate that under the constant-temperature heating mode, after accounting for the thermal delay characteristics of the reactor pool, the number of power adjustments required during the 139-day heating season is reduced by 22 instances. Under the variable-temperature heating mode, the independent heating configuration of this reactor can satisfy the thermal demand during the heating period in cold regions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112162"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146074821","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":"2026-06-01","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}
Pub Date : 2026-06-01Epub Date: 2026-01-21DOI: 10.1016/j.anucene.2026.112157
Ruixiao Zhang , Yuhang Niu , Yanan He , Zhiwei Lu , Yingwei Wu , Jing Zhang , G.H. Su
Silicon carbide (SiC) composite cladding is a promising candidate for accident-tolerant fuel applications, exhibiting distinct thermomechanical and failure behaviors compared to metallic cladding such as Zircaloy under spent fuel storage conditions. Existing safety criteria for spent fuel storage may therefore be inapplicable to SiC-based assemblies. In this study, the updated FROBA code was used to simulate the performance of high-burnup SiC cladding spent fuel after reactor operation, considering long-term wet and dry storage as well as short-term off-normal dry storage. Results show that SiC cladding performs well during wet storage. During dry storage, the cladding stress slightly exceeds the 90 MPa reference limit for Zr cladding. Due to the probabilistic failure characteristics of monolithic SiC, this corresponds to an estimated failure probability of approximately 0.3%. Elevated internal rod pressure is the main contributor to this risk. The cladding peak temperature of 400℃ at the rod top indicates the highest failure risk. Lower storage temperature limits and optimized pressure balance can effectively mitigate failure.
{"title":"Performance assessment of spent fuel storage with duplex SiC composite cladding in Pressurized water reactors","authors":"Ruixiao Zhang , Yuhang Niu , Yanan He , Zhiwei Lu , Yingwei Wu , Jing Zhang , G.H. Su","doi":"10.1016/j.anucene.2026.112157","DOIUrl":"10.1016/j.anucene.2026.112157","url":null,"abstract":"<div><div>Silicon carbide (SiC) composite cladding is a promising candidate for accident-tolerant fuel applications, exhibiting distinct thermomechanical and failure behaviors compared to metallic cladding such as Zircaloy under spent fuel storage conditions. Existing safety criteria for spent fuel storage may therefore be inapplicable to SiC-based assemblies. In this study, the updated FROBA code was used to simulate the performance of high-burnup SiC cladding spent fuel after reactor operation, considering long-term wet and dry storage as well as short-term off-normal dry storage. Results show that SiC cladding performs well during wet storage. During dry storage, the cladding stress slightly exceeds the 90 MPa reference limit for Zr cladding. Due to the probabilistic failure characteristics of monolithic SiC, this corresponds to an estimated failure probability of approximately 0.3%. Elevated internal rod pressure is the main contributor to this risk. The cladding peak temperature of 400℃ at the rod top indicates the highest failure risk. Lower storage temperature limits and optimized pressure balance can effectively mitigate failure.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112157"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035532","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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