Pub Date : 2026-08-01Epub Date: 2026-03-06DOI: 10.1016/j.anucene.2026.112259
Shu Zheng , Daogang Lu , Qiong Cao , Yuxiong Xue
Central measuring shroud is used to provide channels for control rods and protect in-core instrumentation. Positioned above the core outlet, it is subjected to severe thermal shock during SCRAM accidents. A protective cladding is designed outside the central measuring shroud to mitigate thermal shock damage. A no-gap configuration was adopted between the two structures to facilitate engineering assembly. Cladding design was conducted using elastic and elastoplastic constitutive models respectively. The results demonstrate the elastic model fails to yield a cladding thickness meeting all integrity requirements. In contrast, the elastoplastic model identifies a minimum required cladding thickness of 9 mm. Contrary to simplified analytical estimations, increasing cladding thickness reduced overall stress, attributed to enhanced moment of inertia reducing bending stresses induced by interfacial contact moments. Furthermore, the plastic strain can reach or exceed 70% of the elastic strain. Employing an elastoplastic constitutive model is essential for obtaining accurate structural responses and rational design outcomes.
{"title":"Comparative elastoplastic and elastic analysis for the design on no-gap cladding of central measuring shroud against thermal shock","authors":"Shu Zheng , Daogang Lu , Qiong Cao , Yuxiong Xue","doi":"10.1016/j.anucene.2026.112259","DOIUrl":"10.1016/j.anucene.2026.112259","url":null,"abstract":"<div><div>Central measuring shroud is used to provide channels for control rods and protect in-core instrumentation. Positioned above the core outlet, it is subjected to severe thermal shock during SCRAM accidents. A protective cladding is designed outside the central measuring shroud to mitigate thermal shock damage. A no-gap configuration was adopted between the two structures to facilitate engineering assembly. Cladding design was conducted using elastic and elastoplastic constitutive models respectively. The results demonstrate the elastic model fails to yield a cladding thickness meeting all integrity requirements. In contrast, the elastoplastic model identifies a minimum required cladding thickness of 9 mm. Contrary to simplified analytical estimations, increasing cladding thickness reduced overall stress, attributed to enhanced moment of inertia reducing bending stresses induced by interfacial contact moments. Furthermore, the plastic strain can reach or exceed 70% of the elastic strain. Employing an elastoplastic constitutive model is essential for obtaining accurate structural responses and rational design outcomes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112259"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387806","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 : 2026-08-01Epub Date: 2026-03-11DOI: 10.1016/j.anucene.2026.112262
Pingwen Ou , Peng Chen , Yong Ouyang , Chao Guo , Meilan Chen , Dongyu He , Yongzheng Chen
The aerosol behavior plays an important role in the assessment of the source term during severe accidents in nuclear power plants. This paper presents the development and validation of the aerosol behavior model within the source term module of the SPRUCE code, an integral severe accident analysis code independently developed by the China Nuclear Power Technology Research Institute (CNPRI). The model comprehensively incorporates the key physical processes governing aerosol behavior, including deposition (gravitational settling, thermophoresis, diffusiophoresis, and Brownian diffusion), coagulation (Brownian, gravitational, and turbulent), hygroscopic growth, and resuspension. To validate the reliability of model, simulations were compared against the internationally recognized experimental data, including the LACE LA4, KAEVER (K123, K148, K186, K188), and STORM SR11. The results demonstrate that the SPRUCE code successfully predicts the temporal evolution of aerosol concentration and deposition/resuspension behavior in various thermal–hydraulic conditions and aerosol types (soluble, insoluble, and mixed). The favorable agreement with experimental data confirms the capability and credibility of the SPRUCE code and its implemented aerosol models for simulating aerosol behavior under severe accident conditions. This work establishes a solid foundation for the application of SPRUCE in severe accident source term analysis, paving the way for future validation of other critical source term phenomena and other modules of SPRUCE.
{"title":"Aerosol behavior model development and validation of SPRUCE code source term module","authors":"Pingwen Ou , Peng Chen , Yong Ouyang , Chao Guo , Meilan Chen , Dongyu He , Yongzheng Chen","doi":"10.1016/j.anucene.2026.112262","DOIUrl":"10.1016/j.anucene.2026.112262","url":null,"abstract":"<div><div>The aerosol behavior plays an important role in the assessment of the source term during severe accidents in nuclear power plants. This paper presents the development and validation of the aerosol behavior model within the source term module of the SPRUCE code, an integral severe accident analysis code independently developed by the China Nuclear Power Technology Research Institute (CNPRI). The model comprehensively incorporates the key physical processes governing aerosol behavior, including deposition (gravitational settling, thermophoresis, diffusiophoresis, and Brownian diffusion), coagulation (Brownian, gravitational, and turbulent), hygroscopic growth, and resuspension. To validate the reliability of model, simulations were compared against the internationally recognized experimental data, including the LACE LA4, KAEVER (K123, K148, K186, K188), and STORM SR11. The results demonstrate that the SPRUCE code successfully predicts the temporal evolution of aerosol concentration and deposition/resuspension behavior in various thermal–hydraulic conditions and aerosol types (soluble, insoluble, and mixed). The favorable agreement with experimental data confirms the capability and credibility of the SPRUCE code and its implemented aerosol models for simulating aerosol behavior under severe accident conditions. This work establishes a solid foundation for the application of SPRUCE in severe accident source term analysis, paving the way for future validation of other critical source term phenomena and other modules of SPRUCE.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112262"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387855","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 : 2026-08-01Epub Date: 2026-03-11DOI: 10.1016/j.anucene.2026.112269
Yang Yang , Yang Zou
Fluids with internal heat sources exhibit distinct heat transfer characteristics from those without. Using the correlations developed for the latter to predict the hydrodynamic and thermal entrance lengths of molten salt with internal heat source may result in non-negligible errors. Thus, these entrance lengths for laminar molten salt with internal heat source are evaluated using Fluent, with the influences of mass flow rate, inlet temperature, volumetric power density and tube diameter discussed. Results indicate that as the Reynolds number and tube diameter increase, both entrance lengths increase. In contrast, increasing inlet temperature and volumetric power density only increase the hydrodynamic entrance length, while the thermal entrance length remains unchanged. Unlike fluids without internal heat sources, the outlet-to-inlet viscosity ratio exerts an important influence on the hydrodynamic entrance length, rendering existing correlations developed for fluids without internal heat sources invalid. However, part of the thermal entrance length correlations remains accurate within an acceptable tolerance range. Finally, new hydrodynamic and thermal entrance length correlations for laminar molten salt with internal heat source are proposed, with the maximum relative deviations of 9.45% and 1.71% from the numerical results, respectively.
{"title":"Hydrodynamic and thermal entrance lengths for laminar forced convection of molten salt with internal heat source","authors":"Yang Yang , Yang Zou","doi":"10.1016/j.anucene.2026.112269","DOIUrl":"10.1016/j.anucene.2026.112269","url":null,"abstract":"<div><div>Fluids with internal heat sources exhibit distinct heat transfer characteristics from those without. Using the correlations developed for the latter to predict the hydrodynamic and thermal entrance lengths of molten salt with internal heat source may result in non-negligible errors. Thus, these entrance lengths for laminar molten salt with internal heat source are evaluated using Fluent, with the influences of mass flow rate, inlet temperature, volumetric power density and tube diameter discussed. Results indicate that as the Reynolds number and tube diameter increase, both entrance lengths increase. In contrast, increasing inlet temperature and volumetric power density only increase the hydrodynamic entrance length, while the thermal entrance length remains unchanged. Unlike fluids without internal heat sources, the outlet-to-inlet viscosity ratio exerts an important influence on the hydrodynamic entrance length, rendering existing correlations developed for fluids without internal heat sources invalid. However, part of the thermal entrance length correlations remains accurate within an acceptable tolerance range. Finally, new hydrodynamic and thermal entrance length correlations for laminar molten salt with internal heat source are proposed, with the maximum relative deviations of 9.45% and 1.71% from the numerical results, respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112269"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387808","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 : 2026-08-01Epub Date: 2026-03-11DOI: 10.1016/j.anucene.2026.112274
Yunzhi Chai, Qikun Sun, Jiashuang Wan, Shifa Wu
In nuclear power plant system simulations, simulation model accuracy directly influences dynamic characteristic analysis and control system design. To meet real-time requirements, system-level models typically employ simplified modeling approaches, whose internal structural parameters or physical property parameters deviate from the actual system, thereby reducing model accuracy. Furthermore, traditional parameter identification methods often struggle to effectively address time-varying parameters, particularly when operational data is sparse and power range coverage is incomplete. Therefore, this paper proposes an adaptive physics-informed cascaded neural network (PICNN) method for identifying physical property parameters that varies with reactor operating state. The validation results show that, under both noise-free and noisy data conditions, the proposed method has good parameter identification performance and robustness. Moreover, the model output optimized through parameter identification agrees well with the simulation model output data.
{"title":"Adaptive physics-informed cascaded neural networks for nuclear reactor core parameter identification","authors":"Yunzhi Chai, Qikun Sun, Jiashuang Wan, Shifa Wu","doi":"10.1016/j.anucene.2026.112274","DOIUrl":"10.1016/j.anucene.2026.112274","url":null,"abstract":"<div><div>In nuclear power plant system simulations, simulation model accuracy directly influences dynamic characteristic analysis and control system design. To meet real-time requirements, system-level models typically employ simplified modeling approaches, whose internal structural parameters or physical property parameters deviate from the actual system, thereby reducing model accuracy. Furthermore, traditional parameter identification methods often struggle to effectively address time-varying parameters, particularly when operational data is sparse and power range coverage is incomplete. Therefore, this paper proposes an adaptive physics-informed cascaded neural network (PICNN) method for identifying physical property parameters that varies with reactor operating state. The validation results show that, under both noise-free and noisy data conditions, the proposed method has good parameter identification performance and robustness. Moreover, the model output optimized through parameter identification agrees well with the simulation model output data.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112274"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387809","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 : 2026-08-01Epub Date: 2026-03-06DOI: 10.1016/j.anucene.2026.112256
Yasemin Inc, Nithin Puthiyaveettil, Mohammed Siddig
This study evaluates the applicability of Non-Destructive Examination (NDE) methods to the BWRX-300, i-SMR, and Rolls-Royce SMR Small Modular Reactor (SMR) designs, as well as to Generation IV concepts: GTHTR300, IMSR 400, and 4S. The assessment considers plant design, structural materials, and manufacturing technologies to identify corresponding NDE requirements for both manufacturing and in-service inspections. The findings suggest that, for near-term deployable SMRs, existing inspection methods used in the conventional large-scale nuclear power plants remain applicable. However, the introduction of new materials, novel reactor designs, advanced manufacturing techniques, and the modular design, may require adaptations or development of new NDE approaches.
{"title":"Evaluation of non-destructive examination requirements and challenges for small modular reactors","authors":"Yasemin Inc, Nithin Puthiyaveettil, Mohammed Siddig","doi":"10.1016/j.anucene.2026.112256","DOIUrl":"10.1016/j.anucene.2026.112256","url":null,"abstract":"<div><div>This study evaluates the applicability of Non-Destructive Examination (NDE) methods to the BWRX-300, i-SMR, and Rolls-Royce SMR Small Modular Reactor (SMR) designs, as well as to Generation IV concepts: GTHTR300, IMSR 400, and 4S. The assessment considers plant design, structural materials, and manufacturing technologies to identify corresponding NDE requirements for both manufacturing and in-service inspections. The findings suggest that, for near-term deployable SMRs, existing inspection methods used in the conventional large-scale nuclear power plants remain applicable. However, the introduction of new materials, novel reactor designs, advanced manufacturing techniques, and the modular design, may require adaptations or development of new NDE approaches.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112256"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387853","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 : 2026-08-01Epub Date: 2026-03-10DOI: 10.1016/j.anucene.2026.112279
A. Natarajan , N. Mohankumar
We indicate the novel aspects and the practical utility of a new formulation of the transport equation for neutron transport introduced by D.V. Gopinath. In this formulation, the governing equations for the classical Milne problem and the Critical Slab problem are derived by him in a very straightforward way without invoking tools like the Green’s function, Placzek Lemma, etc. In particular, we analyse the Milne and Critical Slab problems under this new method and demonstrate its usefulness for both numerical evaluation and analytical estimates. This analysis is very simple and rigorous and it does not involve any complicated theoretical machinery like the solution of singular integral equations or the Wiener–Hopf method. Interestingly, this approach yields simple but useful approximate analytical estimate for the critical thickness which is believed to be new.
{"title":"An analysis of a new transport equation of Gopinath","authors":"A. Natarajan , N. Mohankumar","doi":"10.1016/j.anucene.2026.112279","DOIUrl":"10.1016/j.anucene.2026.112279","url":null,"abstract":"<div><div>We indicate the novel aspects and the practical utility of a new formulation of the transport equation for neutron transport introduced by D.V. Gopinath. In this formulation, the governing equations for the classical Milne problem and the Critical Slab problem are derived by him in a very straightforward way without invoking tools like the Green’s function, Placzek Lemma, etc. In particular, we analyse the Milne and Critical Slab problems under this new method and demonstrate its usefulness for both numerical evaluation and analytical estimates. This analysis is very simple and rigorous and it does not involve any complicated theoretical machinery like the solution of singular integral equations or the Wiener–Hopf method. Interestingly, this approach yields simple but useful approximate analytical estimate for the critical thickness which is believed to be new.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112279"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387856","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 : 2026-08-01Epub Date: 2026-03-09DOI: 10.1016/j.anucene.2026.112271
Bingyao Zhao , Jie Qiu , Chongdou Yang , Jingjing Liao , Tianyu Zhao , Wei Zhang , Jianqiao Yang , Di Yun
The oxidation behavior of Zr-1.0Sn-1.0Nb alloy in both steam and oxygen atmosphere was studied from 600 ℃ to 1100 ℃ using a simultaneous thermogravimetric analyzer. The results show that a clear, definite and systematic oxidation transitions occurred within the experimental duration. In both steam and oxygen atmospheres, the oxidation transition duration of N36 decreased with increasing temperature, highlighting the significant influence of temperature on the oxidation process. Over a wide temperature range up to 1050 ℃, the oxidation transition time was found to follow an Arrhenius relationship with temperature. The oxidation of N36 is more severe in the oxygen atmosphere than that in the steam atmosphere, as evidenced by a shorter oxidation transition time and a thicker oxide film at the same condition. However, the similar activation energies suggest a common in both atmospheres. By comparing the exponent (n) in oxygen and steam atmospheres, the exponent (n) in oxygen is smaller than that of steam, proving that oxide scale in oxygen is more prone to breakaway, leading to an earlier transition or faster oxidation kinetics in the oxygen atmosphere.
{"title":"Temperature-Dependent oxidation of Zr-Sn-Nb zirconium alloys in LOCA conditions at Elevated temperatures","authors":"Bingyao Zhao , Jie Qiu , Chongdou Yang , Jingjing Liao , Tianyu Zhao , Wei Zhang , Jianqiao Yang , Di Yun","doi":"10.1016/j.anucene.2026.112271","DOIUrl":"10.1016/j.anucene.2026.112271","url":null,"abstract":"<div><div>The oxidation behavior of Zr-1.0Sn-1.0Nb alloy in both steam and oxygen atmosphere was studied from 600 ℃ to 1100 ℃ using a simultaneous thermogravimetric analyzer. The results show that a clear, definite and systematic oxidation transitions occurred within the experimental duration. In both steam and oxygen atmospheres, the oxidation transition duration of N36 decreased with increasing temperature, highlighting the significant influence of temperature on the oxidation process. Over a wide temperature range up to 1050 ℃, the oxidation transition time was found to follow an Arrhenius relationship with temperature. The oxidation of N36 is more severe in the oxygen atmosphere than that in the steam atmosphere, as evidenced by a shorter oxidation transition time and a thicker oxide film at the same condition. However, the similar activation energies suggest a common in both atmospheres. By comparing the exponent (<em>n</em>) in oxygen and steam atmospheres, the exponent (<em>n</em>) in oxygen is smaller than that of steam, proving that oxide scale in oxygen is more prone to breakaway, leading to an earlier transition or faster oxidation kinetics in the oxygen atmosphere.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112271"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387858","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}
DLOFC accidents in HTGRs present potential radiological hazards to both human health and the environment. However, existing source term models often rely on simplified assumptions, introducing significant uncertainties in accident consequence assessments. To enhance simulation reliability, this study develops a multiscale DLOFC source term analysis framework that captures the accident’s physical processes across temporal and spatial dimensions. The framework integrates a suite of high-fidelity, interconnected models to quantitatively characterize radionuclide transport and distribution throughout all key accident phases and reactor regions. The core source term is simulated using a multiphysics-coupled model, while a statistical ensemble of fuel pebble operational histories under steady-state conditions defines the initial accident state, enabling full-core release analysis throughout the accident progression. Radionuclide distribution within the nuclear island prior to depressurization is determined via a one-dimensional, four-zone transport-and-plateout model, combined with a detailed nodalization of the primary loop. Additionally, the framework incorporates the FRG desorption model alongside operational data to simulate radionuclide migration across the core, primary loop, containment, and environment throughout different accident phases. Using the HTR-PM as a reference case, a comprehensive DLOFC source term analysis is conducted. The results indicate that the environmental release fractions for typical nuclides are for the long-lived inert gas Kr-85, for the short-lived inert gas Xe-133, and for the long-lived metallic nuclide Cs-137. Compared to conventional accident analysis models, the radiological releases predicted by the present framework are approximately an order of magnitude lower, which can be attributed to the adoption of a more realistic core fission product transport-release model and a more reasonable fuel performance analysis model. These findings demonstrate that the proposed framework enhances the completeness and accuracy of source term assessment, providing robust technical support for underscoring the inherent safety features of HTGRs.
{"title":"Development of a multiscale DLOFC source term analysis framework for pebble bed HTGR","authors":"Chenghao Cao, Junyi Chen, Shaoning Shen, Jingang Liang, Chuan Li, Jianzhu Cao","doi":"10.1016/j.anucene.2026.112261","DOIUrl":"10.1016/j.anucene.2026.112261","url":null,"abstract":"<div><div>DLOFC accidents in HTGRs present potential radiological hazards to both human health and the environment. However, existing source term models often rely on simplified assumptions, introducing significant uncertainties in accident consequence assessments. To enhance simulation reliability, this study develops a multiscale DLOFC source term analysis framework that captures the accident’s physical processes across temporal and spatial dimensions. The framework integrates a suite of high-fidelity, interconnected models to quantitatively characterize radionuclide transport and distribution throughout all key accident phases and reactor regions. The core source term is simulated using a multiphysics-coupled model, while a statistical ensemble of fuel pebble operational histories under steady-state conditions defines the initial accident state, enabling full-core release analysis throughout the accident progression. Radionuclide distribution within the nuclear island prior to depressurization is determined via a one-dimensional, four-zone transport-and-plateout model, combined with a detailed nodalization of the primary loop. Additionally, the framework incorporates the FRG desorption model alongside operational data to simulate radionuclide migration across the core, primary loop, containment, and environment throughout different accident phases. Using the HTR-PM as a reference case, a comprehensive DLOFC source term analysis is conducted. The results indicate that the environmental release fractions for typical nuclides are <span><math><mrow><mn>3</mn><mo>.</mo><mn>2</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mo>−</mo><mn>8</mn></mrow></msup></mrow></math></span> for the long-lived inert gas Kr-85, <span><math><mrow><mn>1</mn><mo>.</mo><mn>4</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mo>−</mo><mn>7</mn></mrow></msup></mrow></math></span> for the short-lived inert gas Xe-133, and <span><math><mrow><mn>6</mn><mo>.</mo><mn>7</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mo>−</mo><mn>8</mn></mrow></msup></mrow></math></span> for the long-lived metallic nuclide Cs-137. Compared to conventional accident analysis models, the radiological releases predicted by the present framework are approximately an order of magnitude lower, which can be attributed to the adoption of a more realistic core fission product transport-release model and a more reasonable fuel performance analysis model. These findings demonstrate that the proposed framework enhances the completeness and accuracy of source term assessment, providing robust technical support for underscoring the inherent safety features of HTGRs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112261"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147388276","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 : 2026-08-01Epub Date: 2026-03-10DOI: 10.1016/j.anucene.2026.112260
Hongjia Liu , Xiong Zhang , Shuang Zhang , Peng Li , Wei Liu , Linhe Du , Yanbin Wang , Yunfei Zhu , Nan Chao , Xiaoqiang Li
Nuclear radiation propagation pathways and dose distributions in urban environments are influenced by multiple factors, including the construction materials and the spatial distribution characteristics of buildings within urban scenarios, which render the radiation transport problem and spatial radiation distribution highly intricate. Rapid construction of quasi-realistic urban building scenarios constitutes a prerequisite for conducting Monte Carlo simulations to investigate radiation transport mechanisms and dose distribution patterns in urban environments. This study develops an automated modeling methodology for global urban buildings based on Constructive Solid Geometry (CSG), leveraging open-source global Geographic Information System (GIS) urban databases to rapidly reconstruct buildings without building feature information. This provides crucial technical support for investigating the effects of urban buildings on radiation and their influence on altering radiation field spatial distributions through scattering phenomena, and the impacts of building spatial distribution patterns and material compositions on radiation transmission and radiation field distribution characteristics in urban environments.
{"title":"Automated building modeling methodology for Monte Carlo simulation based on global GIS urban datasets","authors":"Hongjia Liu , Xiong Zhang , Shuang Zhang , Peng Li , Wei Liu , Linhe Du , Yanbin Wang , Yunfei Zhu , Nan Chao , Xiaoqiang Li","doi":"10.1016/j.anucene.2026.112260","DOIUrl":"10.1016/j.anucene.2026.112260","url":null,"abstract":"<div><div>Nuclear radiation propagation pathways and dose distributions in urban environments are influenced by multiple factors, including the construction materials and the spatial distribution characteristics of buildings within urban scenarios, which render the radiation transport problem and spatial radiation distribution highly intricate. Rapid construction of quasi-realistic urban building scenarios constitutes a prerequisite for conducting Monte Carlo simulations to investigate radiation transport mechanisms and dose distribution patterns in urban environments. This study develops an automated modeling methodology for global urban buildings based on Constructive Solid Geometry (CSG), leveraging open-source global Geographic Information System (GIS) urban databases to rapidly reconstruct buildings without building feature information. This provides crucial technical support for investigating the effects of urban buildings on radiation and their influence on altering radiation field spatial distributions through scattering phenomena, and the impacts of building spatial distribution patterns and material compositions on radiation transmission and radiation field distribution characteristics in urban environments.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112260"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387811","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 : 2026-08-01Epub Date: 2026-03-05DOI: 10.1016/j.anucene.2026.112263
Ian T. Kolaja , Lee A. Bernstein , Ludovic Jantzen , Eleanor Tubman , Tatiana Siaraferas , Massimiliano Fratoni
Burnup measurement is essential for monitoring and operating pebble-bed reactors (PBRs), where fuel pebbles circulate rapidly through the core. However, conventional gamma spectroscopy using high-purity germanium (HPGe) detectors is challenging due to high activity levels in discharge pebbles, leading to excessive dead time and Compton scattering. This study explores the use of bent crystal diffraction (BCD) spectrometers to filter the emitted gamma spectrum and isolate key peaks for improved measurement accuracy and speed. Pebble-wise depletion calculations were performed and the resulting spectra were analyzed using ray tracing (SHADOW3) and gamma response modeling (GADRAS). Key isotopes, Ba/137Cs, 239Np, 144Ce, Pm, and 140La, were found to strongly correlate with burnup, residence time, core passes, plutonium production, and fluence. Machine learning regression models that were given synthetic spectra achieved a coefficient of determination () as high as 0.995 for burnup prediction. Among various BCD configurations, mosaic silicon crystals in the (440) orientation combined with an HPGe detector provided optimal performance for measuring 137Cs decay (via Ba), while silicon (220) and (440) paired with scintillators were effective for the shorter-lived isotopes.
{"title":"Burnup measurement using bent crystal diffraction spectrometers for pebble bed reactors","authors":"Ian T. Kolaja , Lee A. Bernstein , Ludovic Jantzen , Eleanor Tubman , Tatiana Siaraferas , Massimiliano Fratoni","doi":"10.1016/j.anucene.2026.112263","DOIUrl":"10.1016/j.anucene.2026.112263","url":null,"abstract":"<div><div>Burnup measurement is essential for monitoring and operating pebble-bed reactors (PBRs), where fuel pebbles circulate rapidly through the core. However, conventional gamma spectroscopy using high-purity germanium (HPGe) detectors is challenging due to high activity levels in discharge pebbles, leading to excessive dead time and Compton scattering. This study explores the use of bent crystal diffraction (BCD) spectrometers to filter the emitted gamma spectrum and isolate key peaks for improved measurement accuracy and speed. Pebble-wise depletion calculations were performed and the resulting spectra were analyzed using ray tracing (SHADOW3) and gamma response modeling (GADRAS). Key isotopes, <span><math><msup><mrow></mrow><mrow><mn>137</mn><mi>m</mi></mrow></msup></math></span>Ba/<sup>137</sup>Cs, <sup>239</sup>Np, <sup>144</sup>Ce, <span><math><msup><mrow></mrow><mrow><mn>148</mn><mi>m</mi></mrow></msup></math></span>Pm, and <sup>140</sup>La, were found to strongly correlate with burnup, residence time, core passes, plutonium production, and fluence. Machine learning regression models that were given synthetic spectra achieved a coefficient of determination (<span><math><msup><mrow><mi>R</mi></mrow><mrow><mn>2</mn></mrow></msup></math></span>) as high as 0.995 for burnup prediction. Among various BCD configurations, mosaic silicon crystals in the (440) orientation combined with an HPGe detector provided optimal performance for measuring <sup>137</sup>Cs decay (via <span><math><msup><mrow></mrow><mrow><mn>137</mn><mi>m</mi></mrow></msup></math></span>Ba), while silicon (220) and (440) paired with scintillators were effective for the shorter-lived isotopes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112263"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387852","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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