Pub Date : 2025-02-11DOI: 10.1016/j.anucene.2025.111245
Qingfeng Jiang, Pengfei Wang
The megawatt novel nuclear power system (MNNPS) employing heat-pipe reactors and sCO2 Brayton cycles has significant applications. However, the unattended operating environment places high demands on the adaptive capability. In this paper, a fuzzy logic-based coordinated adaptive control (FL-CAC) method for MNNPS is proposed. The key to the method is to design fuzzy logic controllers with the current loop-controlled variable as the main-input and the remaining loop-controlled variables as the auxiliary-inputs, and to realize the coordinated adaptive adjustment of control parameters by comprehensively considering the operating status in control loops. The FL-CAC method was verified under typical operating conditions such as step and ramp load change transients. The simulation results demonstrates that it can significantly improve the control performance compared with the original control scheme. This study can provide a theoretical reference for the autonomous and reliable control of MNNPS which need to operate in an unattended environment.
{"title":"A fuzzy logic-based coordinated adaptive control method for megawatt novel nuclear power systems","authors":"Qingfeng Jiang, Pengfei Wang","doi":"10.1016/j.anucene.2025.111245","DOIUrl":"10.1016/j.anucene.2025.111245","url":null,"abstract":"<div><div>The megawatt novel nuclear power system (MNNPS) employing heat-pipe reactors and sCO<sub>2</sub> Brayton cycles has significant applications. However, the unattended operating environment places high demands on the adaptive capability. In this paper, a fuzzy logic-based coordinated adaptive control (FL-CAC) method for MNNPS is proposed. The key to the method is to design fuzzy logic controllers with the current loop-controlled variable as the main-input and the remaining loop-controlled variables as the auxiliary-inputs, and to realize the coordinated adaptive adjustment of control parameters by comprehensively considering the operating status in control loops. The FL-CAC method was verified under typical operating conditions such as step and ramp load change transients. The simulation results demonstrates that it can significantly improve the control performance compared with the original control scheme. This study can provide a theoretical reference for the autonomous and reliable control of MNNPS which need to operate in an unattended environment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111245"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387333","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-11DOI: 10.1016/j.anucene.2025.111258
Jeremy Bittan , Nikolai Bakouta , Julie-Anne Zambaux , Laure Carénini
This paper presents a comparison between EDF MAAP 5.04 and ASTECv3 codes for a hypothetical severe accident leading to core degradation in the ELSMOR project’s proposed E-SMR (Small Modular Reactor) design. The ELSMOR (Towards European Licensing of Small Modular Reactors) project was a Horizon 2020 Euratom project that ended in 2023. The consortium included 15 partners from 8 European countries, involving research institutes, major European nuclear companies, and technical support organizations. The 3.5-year project, launched in September 2019, investigated selected safety features of Light-Water (LW) SMRs with a focus on licensing aspects.
The Modular Accident Analysis Program (MAAP) is a deterministic code owned and licensed by the Electric Power Research Institute (EPRI) that simulates the response of light water moderated nuclear power plants during accidental transients. ASTEC v3, developed by IRSN, is a severe accident system code used to evaluate major nuclear accidents for different nuclear installations, focusing on western light water reactor designs.
This study compares EDF MAAP5.04 and ASTECv3 in terms of transient evolution from the initiating event (a Station Black Out), core degradation and hydrogen generation, corium relocation to the lower plenum, and in-vessel melt retention (IVMR). The comparison also includes physical phenomena in the containment, such as steam condensation on the walls. The potential for H2 combustion, based on the specific assumptions of the selected transient, is evaluated through flammability diagrams, and a sensitivity analysis of N2 injection for inerting the containment is assessed.
This study shows that both EDF MAAP 5.04 and ASTECv3 codes can effectively simulate severe accident phenomena and mitigation strategies like IVMR, containment inerting, and reactor decay heat removal. The codes generally agree well, providing key parameter magnitudes. Some modeling improvements were made, and remaining discrepancies are discussed. Sensitivity calculations are presented to confirm the analysis and to highlight model uncertainties, particularly for steam condensation and hydrogen production.
{"title":"Comparison between EDF MAAP5.04 and ASTECV3 codes on a hypothetical severe accident on the ELSMOR project E-SMR design","authors":"Jeremy Bittan , Nikolai Bakouta , Julie-Anne Zambaux , Laure Carénini","doi":"10.1016/j.anucene.2025.111258","DOIUrl":"10.1016/j.anucene.2025.111258","url":null,"abstract":"<div><div>This paper presents a comparison between EDF MAAP 5.04 and ASTECv3 codes for a hypothetical severe accident leading to core degradation in the ELSMOR project’s proposed E-SMR (Small Modular Reactor) design. The ELSMOR (Towards European Licensing of Small Modular Reactors) project was a Horizon 2020 Euratom project that ended in 2023. The consortium included 15 partners from 8 European countries, involving research institutes, major European nuclear companies, and technical support organizations. The 3.5-year project, launched in September 2019, investigated selected safety features of Light-Water (LW) SMRs with a focus on licensing aspects.</div><div>The Modular Accident Analysis Program (MAAP) is a deterministic code owned and licensed by the Electric Power Research Institute (EPRI) that simulates the response of light water moderated nuclear power plants during accidental transients. ASTEC v3, developed by IRSN, is a severe accident system code used to evaluate major nuclear accidents for different nuclear installations, focusing on western light water reactor designs.</div><div>This study compares EDF MAAP5.04 and ASTECv3 in terms of transient evolution from the initiating event (a Station Black Out), core degradation and hydrogen generation, corium relocation to the lower plenum, and in-vessel melt retention (IVMR). The comparison also includes physical phenomena in the containment, such as steam condensation on the walls. The potential for H<sub>2</sub> combustion, based on the specific assumptions of the selected transient, is evaluated through flammability diagrams, and a sensitivity analysis of N2 injection for inerting the containment is assessed.</div><div>This study shows that both EDF MAAP 5.04 and ASTECv3 codes can effectively simulate severe accident phenomena and mitigation strategies like IVMR, containment inerting, and reactor decay heat removal. The codes generally agree well, providing key parameter magnitudes. Some modeling improvements were made, and remaining discrepancies are discussed. Sensitivity calculations are presented to confirm the analysis and to highlight model uncertainties, particularly for steam condensation and hydrogen production.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111258"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387331","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-11DOI: 10.1016/j.anucene.2025.111235
P. Cosgrove, N. Read
It was recently argued by Kemp et al. that HALEU (high-assay low-enriched uranium, or uranium enriched up to 19.75%) can conceivably be used to produce a nuclear weapon and on this basis civilian enrichment limits should be lowered to 10% or 12% (Scott Kemp et al., 2024). We find their argument unconvincing in several respects.
{"title":"On the practicalities of producing a nuclear weapon using high-assay low-enriched uranium","authors":"P. Cosgrove, N. Read","doi":"10.1016/j.anucene.2025.111235","DOIUrl":"10.1016/j.anucene.2025.111235","url":null,"abstract":"<div><div>It was recently argued by Kemp et al. that HALEU (high-assay low-enriched uranium, or uranium enriched up to 19.75%) can conceivably be used to produce a nuclear weapon and on this basis civilian enrichment limits should be lowered to 10% or 12% (Scott Kemp et al., 2024). We find their argument unconvincing in several respects.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111235"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143379075","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-10DOI: 10.1016/j.anucene.2025.111236
Dhongik S. Yoon , Firat Cetinbas , Guanyi Wang , John A. Stillman , Valerio Mascolino , Maria Pinilla , Earl E. Feldman , Walid Mohamed , Erik H. Wilson
A series of structural analyses have been performed to support the conversion of the University of Missouri Research Reactor (MURR) from the use of highly enriched uranium (HEU; ≥20 wt% U-235) to low-enriched uranium (LEU; <20 wt% U-235) fuel. The irradiation thermo-mechanical analysis evaluated the effects of fuel swelling, irradiation creep, thermal expansion, as well as thermal resistance from the oxide layer growth for the MURR LEU element in prototypic thermal and irradiation conditions as presented in Part I of this article. Overall, this irradiation thermo-mechanical analysis predicts smaller gap thickness reductions in previously limiting regions, and larger reductions in the middle of the outermost end channels where power density is not typically a maximum. Due to substantial differences between the channel gap reductions assumed for the previous safety analyses and those predicted by the irradiation thermo-mechanical analysis, a need to evaluate their impact on the thermal hydraulics safety analyses arose. This article presents the results from the steady-state safety analysis for normal operation as well as the two most limiting accident scenarios. The calculation models were revised in order to account for the spatial and temporal variation of the channel gap thicknesses. The results show that sufficient safety margins are still maintained for normal operation as well as during the postulated accident transients. This work provides a methodology of incorporating the irradiation thermo-mechanical behavior of plate-type fuel into the thermal hydraulics safety analyses.
{"title":"MURR LEU structural and thermal hydraulics analyses: Part II – Impacts of irradiation thermo-mechanical behavior on thermal hydraulics safety analyses","authors":"Dhongik S. Yoon , Firat Cetinbas , Guanyi Wang , John A. Stillman , Valerio Mascolino , Maria Pinilla , Earl E. Feldman , Walid Mohamed , Erik H. Wilson","doi":"10.1016/j.anucene.2025.111236","DOIUrl":"10.1016/j.anucene.2025.111236","url":null,"abstract":"<div><div>A series of structural analyses have been performed to support the conversion of the University of Missouri Research Reactor (MURR) from the use of highly enriched uranium (HEU; ≥20 wt% U-235) to low-enriched uranium (LEU; <20 wt% U-235) fuel. The irradiation thermo-mechanical analysis evaluated the effects of fuel swelling, irradiation creep, thermal expansion, as well as thermal resistance from the oxide layer growth for the MURR LEU element in prototypic thermal and irradiation conditions as presented in Part I of this article. Overall, this irradiation thermo-mechanical analysis predicts smaller gap thickness reductions in previously limiting regions, and larger reductions in the middle of the outermost end channels where power density is not typically a maximum. Due to substantial differences between the channel gap reductions assumed for the previous safety analyses and those predicted by the irradiation thermo-mechanical analysis, a need to evaluate their impact on the thermal hydraulics safety analyses arose. This article presents the results from the steady-state safety analysis for normal operation as well as the two most limiting accident scenarios. The calculation models were revised in order to account for the spatial and temporal variation of the channel gap thicknesses. The results show that sufficient safety margins are still maintained for normal operation as well as during the postulated accident transients. This work provides a methodology of incorporating the irradiation thermo-mechanical behavior of plate-type fuel into the thermal hydraulics safety analyses.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111236"},"PeriodicalIF":1.9,"publicationDate":"2025-02-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143376550","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-10DOI: 10.1016/j.anucene.2025.111242
Guanwen Luo, Kuo Wang, Gang Tong, Chong Xie, Xiaohang Wu, Huiyong Zhang, Yiliang Xie, Jiayi Chen
The paper performs numerical analysis on a 19-rod bundle LBE cooled fuel assembly, considering non-uniform rods power distribution. Five typical rods power distribution modes are established. The coolant velocity field, temperature distribution characteristics and heat transfer behavior under various modes are explored. The work also compares thermal–hydraulic behaviors of forced circulation with that of natural circulation condition. Results show the rods with high power-rate enhance surrounding coolant flow. Temperature difference is the largest in internal heated mode. When adopting radial increment mode and peripheral heated mode, the PCT is smaller and temperature distributes more evenly. In terms of heat transfer, there are larger local Nu in the subchannels near the high-power rods, and the largest overall Nu occurs in radial increment mode. The thermal–hydraulic characteristics is more susceptible to non-uniform rods heated condition at natural circulation. Eventually, the suggestions for LBE cooled fuel assembly design and operation are proposed.
{"title":"Thermal-hydraulic analysis of a 19-rod bundle LBE cooled fuel assembly with non-uniform rods power distribution by numerical simulation","authors":"Guanwen Luo, Kuo Wang, Gang Tong, Chong Xie, Xiaohang Wu, Huiyong Zhang, Yiliang Xie, Jiayi Chen","doi":"10.1016/j.anucene.2025.111242","DOIUrl":"10.1016/j.anucene.2025.111242","url":null,"abstract":"<div><div>The paper performs numerical analysis on a 19-rod bundle LBE cooled fuel assembly, considering non-uniform rods power distribution. Five typical rods power distribution modes are established. The coolant velocity field, temperature distribution characteristics and heat transfer behavior under various modes are explored. The work also compares thermal–hydraulic behaviors of forced circulation with that of natural circulation condition. Results show the rods with high power-rate enhance surrounding coolant flow. Temperature difference is the largest in internal heated mode. When adopting radial increment mode and peripheral heated mode, the <em>PCT</em> is smaller and temperature distributes more evenly. In terms of heat transfer, there are larger local <em>Nu</em> in the subchannels near the high-power rods, and the largest overall <em>Nu</em> occurs in radial increment mode. The thermal–hydraulic characteristics is more susceptible to non-uniform rods heated condition at natural circulation. Eventually, the suggestions for LBE cooled fuel assembly design and operation are proposed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111242"},"PeriodicalIF":1.9,"publicationDate":"2025-02-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143376548","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}
Heavy water moderated-molten salt reactor (HWMSR) is a novel concept in which heavy water is used as the neutron moderator in lieu of graphite. Excellent neutron economy of heavy water moderator allows HWMSR to use natural uranium as its fuel, foregoing the need for uranium enrichment. Studies regarding HWMSR typically assumes a single fluid stream with identical fuel conduit size throughout the core. This study expands that premise by analysing the performance of HWMSR fuelled by natural uranium with virtual one-and-half fluid stream, using 10% fuel volume fraction (VF) in the narrow fuel conduit and 22.5% fuel VF in blanket conduit. The variation ranges from no blanket until three outer blankets. Neutronic and burnup calculations were conducted using MCNP6.2 code and ENDF/B-VII.0 neutron cross section library. From the calculation results, it was observed that the temperature coefficient of reactivity (TCR) for all variations worsen over time, and only one variation remains negative until the end of cycle (EOC). Fuel cycle length in larger blanket layer numbers is significantly longer, better fertile conversion, and lower fissile consumption. The plutonium vector degraded quickly, so that it becomes impossible to divert. Overall, it is suggested that HWMSR that works using natural uranium be given larger blanket layer numbers to maintain safety and improving performance.
{"title":"Improvement on natural uranium utilisation in heavy water-moderated molten salt reactor using radial blanket","authors":"R.Andika Putra Dwijayanto , Fitria Miftasani , Nina Widiawati , Andang Widi Harto","doi":"10.1016/j.anucene.2025.111257","DOIUrl":"10.1016/j.anucene.2025.111257","url":null,"abstract":"<div><div>Heavy water moderated-molten salt reactor (HWMSR) is a novel concept in which heavy water is used as the neutron moderator in lieu of graphite. Excellent neutron economy of heavy water moderator allows HWMSR to use natural uranium as its fuel, foregoing the need for uranium enrichment. Studies regarding HWMSR typically assumes a single fluid stream with identical fuel conduit size throughout the core. This study expands that premise by analysing the performance of HWMSR fuelled by natural uranium with virtual one-and-half fluid stream, using 10% fuel volume fraction (VF) in the narrow fuel conduit and 22.5% fuel VF in blanket conduit. The variation ranges from no blanket until three outer blankets. Neutronic and burnup calculations were conducted using MCNP6.2 code and ENDF/B-VII.0 neutron cross section library. From the calculation results, it was observed that the temperature coefficient of reactivity (TCR) for all variations worsen over time, and only one variation remains negative until the end of cycle (EOC). Fuel cycle length in larger blanket layer numbers is significantly longer, better fertile conversion, and lower fissile consumption. The plutonium vector degraded quickly, so that it becomes impossible to divert. Overall, it is suggested that HWMSR that works using natural uranium be given larger blanket layer numbers to maintain safety and improving performance.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111257"},"PeriodicalIF":1.9,"publicationDate":"2025-02-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143376549","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-08DOI: 10.1016/j.anucene.2025.111239
Shixiang Hu , Yunzhi Tan , De’an Sun , Keyu Wu
Bentonite pellets are commonly regarded as an excellent sealing/backfilling material for high-level nuclear waste repositories. This study investigated the particle motion characteristics and segregation degree in bentonite pellet mixtures during a vibration process. Results showed that the packing homogeneity of mixtures with multi-size classes was always higher than that of mixtures with binary-size classes during vibration. The mixtures with binary-size classes achieve structural stability more quickly due to the percolation, while the mixtures with multi-size classes remain in motion. Additionally, hydration tests were conducted to analyze the water injection rate and pore-structure distribution evolution in mixtures. Uniformly mixed specimens exhibit faster injected water rates during the hydration process. After long-term hydration, the specimens achieved macroscopic homogeneity, but significant heterogeneity still exists in microstructure. The uniformly mixed method caused higher homogeneity in specimens with binary-size classes, and the layered filling achieved higher homogeneity in specimens with multi-size classes.
{"title":"Investigation on the packing and hydration homogeneity of bentonite pellet mixtures","authors":"Shixiang Hu , Yunzhi Tan , De’an Sun , Keyu Wu","doi":"10.1016/j.anucene.2025.111239","DOIUrl":"10.1016/j.anucene.2025.111239","url":null,"abstract":"<div><div>Bentonite pellets are commonly regarded as an excellent sealing/backfilling material for high-level nuclear waste repositories. This study investigated the particle motion characteristics and segregation degree in bentonite pellet mixtures during a vibration process. Results showed that the packing homogeneity of mixtures with multi-size classes was always higher than that of mixtures with binary-size classes during vibration. The mixtures with binary-size classes achieve structural stability more quickly due to the percolation, while the mixtures with multi-size classes remain in motion. Additionally, hydration tests were conducted to analyze the water injection rate and pore-structure distribution evolution in mixtures. Uniformly mixed specimens exhibit faster injected water rates during the hydration process. After long-term hydration, the specimens achieved macroscopic homogeneity, but significant heterogeneity still exists in microstructure. The uniformly mixed method caused higher homogeneity in specimens with binary-size classes, and the layered filling achieved higher homogeneity in specimens with multi-size classes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111239"},"PeriodicalIF":1.9,"publicationDate":"2025-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143350692","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-08DOI: 10.1016/j.anucene.2025.111248
M.A. Uvakin, A.L. Nikolaev, M.V. Antipov, I.V. Makhin
The work aimed to the development of a methodology for calculating safety assessment of VVER reactor plants in maneuvering modes. This methodology was developed at OKB “GIDROPRESS” to solve the problem of conducting safety analyzes of high-power VVER reactor plants in a flexible operation mode. Performing such work directly requires significant computing resources and multi-parameter expert assessments. Therefore, the main direction of development was the use of a numerical method using neural network models. In particular, the possibility of efficiency increasing of calculations show in terms of choosing the moment in time when the occurrence of the initial event leads to the most conservative results.
In this work, we study the possibilities of further development of the method by constructing neural networks with deep learning aimed at predicting the development of non-stationary processes, taking into account a large number and complex relationships of available parameters. The capabilities of convolution and recursive architectures for constructing neural networks analyzed to estimate the reactor plant dynamics, taking into account maneuvering after an accident occurs. The analysis examines the interpretability of the results in terms of accounting for xenon transients, water exchange operations, and control movement. For software implementation of the method, the VELETMA/GP program is used.
Based on the results of the work, conclusions drawn about the practical significance of the methods used for solving the tasks set for the calculation substantiation of designs of reactor plants with VVER in maneuvering modes. The work uses both the experience of computational justification and the results of validating full-scale tests of maneuvering modes on modern high-power VVER reactors.
{"title":"Artificial intelligence methods application for reactor dynamics predicting in the tasks of maneuverable modes safety assessment","authors":"M.A. Uvakin, A.L. Nikolaev, M.V. Antipov, I.V. Makhin","doi":"10.1016/j.anucene.2025.111248","DOIUrl":"10.1016/j.anucene.2025.111248","url":null,"abstract":"<div><div>The work aimed to the development of a methodology for calculating safety assessment of VVER reactor plants in maneuvering modes. This methodology was developed at OKB “GIDROPRESS” to solve the problem of conducting safety analyzes of high-power VVER reactor plants in a flexible operation mode. Performing such work directly requires significant computing resources and multi-parameter expert assessments. Therefore, the main direction of development was the use of a numerical method using neural network models. In particular, the possibility of efficiency increasing of calculations show in terms of choosing the moment in time when the occurrence of the initial event leads to the most conservative results.</div><div>In this work, we study the possibilities of further development of the method by constructing neural networks with deep learning aimed at predicting the development of non-stationary processes, taking into account a large number and complex relationships of available parameters. The capabilities of convolution and recursive architectures for constructing neural networks analyzed to estimate the reactor plant dynamics, taking into account maneuvering after an accident occurs. The analysis examines the interpretability of the results in terms of accounting for xenon transients, water exchange operations, and control movement. For software implementation of the method, the VELETMA/GP program is used.</div><div>Based on the results of the work, conclusions drawn about the practical significance of the methods used for solving the tasks set for the calculation substantiation of designs of reactor plants with VVER in maneuvering modes. The work uses both the experience of computational justification and the results of validating full-scale tests of maneuvering modes on modern high-power VVER reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111248"},"PeriodicalIF":1.9,"publicationDate":"2025-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143350693","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}
<div><div>Microreactors, characterized by their complex geometries, distinctive neutron spectrum and spatial distributions, pose significant challenges for deterministic nuclear reactor core calculations. Conventional simulation methods, such as assembly homogenization techniques, are commonly employed in microreactor simulations but often fail to fully capture their complexities. To obtain high-resolution simulation, this paper explores two different types of neutronics calculation models based on an explicit description of the reactor core geometry: a high-order model that incorporates multi-group cross-sections and neutron transport calculations, and a low-order model that utilizes few-group cross-sections with the Super-homogenization (SPH) correction and neutron diffusion calculations. Initially, an extensive error analysis is conducted on the low-order model by varying the energy group structure and spatial region division. Subsequently, the high-order and low-order models are compared. The findings indicate that the high-order model achieves superior computational accuracy, with a <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>eff</mi></mrow></msub></math></span> error of −75 pcm, and maximum relative errors and root mean square relative errors (RRMSE) of power at 3.8% and 1.1%, respectively. In contrast, the initial maximum relative power error and RRMSE using the uncorrected low-order model are 10.1% and 2.7%, respectively. After implementing the SPH method, the low-order model reduces these errors to 7.1% and 1.8%, respectively. Moreover, the low-order model demonstrates substantial improvements in memory usage and computational speed, with enhancements of 385-time and 247-time, respectively. Thus, the low-order model offers a balanced approach by enhancing accuracy compared to conventional diffusion calculation model while maintaining efficient resource usage, achieving higher accuracy in power distribution results at a lower computational cost.</div><div>Conventional simulation methods, such as assembly homogenization techniques, are commonly employed in microreactor simulations but often fail to fully capture their complexities. To obtain high-resolution simulation, this paper explores two neutronics calculation models: a high-order model that incorporates multi-group cross-sections and neutron transport calculations, and a low-order model that utilizes few-group cross-sections with the Super-homogenization (SPH) correction and neutron diffusion calculations. The findings indicate that the high-order model achieves superior computational accuracy, with a <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>eff</mi></mrow></msub></math></span> error of −75 pcm, and root mean square relative errors of power at 1.1%. After implementing the SPH method, the low-order model reduces the diffusion errors to 102 pcm and 1.8%, respectively. Moreover, the low-order model improves 385-time memory usage and 247-time computational speed. Thus, t
{"title":"A comparative study of high and low order neutronics calculation models in microreactors","authors":"Xiangyue Li, Qizheng Sun, Xiaojing Liu, Xiang Chai, Hui He, Tengfei Zhang","doi":"10.1016/j.anucene.2025.111205","DOIUrl":"10.1016/j.anucene.2025.111205","url":null,"abstract":"<div><div>Microreactors, characterized by their complex geometries, distinctive neutron spectrum and spatial distributions, pose significant challenges for deterministic nuclear reactor core calculations. Conventional simulation methods, such as assembly homogenization techniques, are commonly employed in microreactor simulations but often fail to fully capture their complexities. To obtain high-resolution simulation, this paper explores two different types of neutronics calculation models based on an explicit description of the reactor core geometry: a high-order model that incorporates multi-group cross-sections and neutron transport calculations, and a low-order model that utilizes few-group cross-sections with the Super-homogenization (SPH) correction and neutron diffusion calculations. Initially, an extensive error analysis is conducted on the low-order model by varying the energy group structure and spatial region division. Subsequently, the high-order and low-order models are compared. The findings indicate that the high-order model achieves superior computational accuracy, with a <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>eff</mi></mrow></msub></math></span> error of −75 pcm, and maximum relative errors and root mean square relative errors (RRMSE) of power at 3.8% and 1.1%, respectively. In contrast, the initial maximum relative power error and RRMSE using the uncorrected low-order model are 10.1% and 2.7%, respectively. After implementing the SPH method, the low-order model reduces these errors to 7.1% and 1.8%, respectively. Moreover, the low-order model demonstrates substantial improvements in memory usage and computational speed, with enhancements of 385-time and 247-time, respectively. Thus, the low-order model offers a balanced approach by enhancing accuracy compared to conventional diffusion calculation model while maintaining efficient resource usage, achieving higher accuracy in power distribution results at a lower computational cost.</div><div>Conventional simulation methods, such as assembly homogenization techniques, are commonly employed in microreactor simulations but often fail to fully capture their complexities. To obtain high-resolution simulation, this paper explores two neutronics calculation models: a high-order model that incorporates multi-group cross-sections and neutron transport calculations, and a low-order model that utilizes few-group cross-sections with the Super-homogenization (SPH) correction and neutron diffusion calculations. The findings indicate that the high-order model achieves superior computational accuracy, with a <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>eff</mi></mrow></msub></math></span> error of −75 pcm, and root mean square relative errors of power at 1.1%. After implementing the SPH method, the low-order model reduces the diffusion errors to 102 pcm and 1.8%, respectively. Moreover, the low-order model improves 385-time memory usage and 247-time computational speed. Thus, t","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111205"},"PeriodicalIF":1.9,"publicationDate":"2025-02-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143348299","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-07DOI: 10.1016/j.anucene.2025.111246
Sandro Pelloni, Dimitri Rochman
<div><div>A general method is proposed to choose, among several benchmarks, suitable assimilation database parameters to be used in adjustments. All these benchmarks are predetermined keeping in mind specific goals of the data assimilation. The main principle of the novel methodology is that of excluding groups of strongly cross-correlated computed parameters by estimating their posterior values of <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span>. The assimilation tries minimizing the mean <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span> for the entirety of the considered benchmarks, including those target parameters which are not part of the selected assimilation database.</div><div>The methodology currently applied to the JEFF-3.3 library, is demonstrated by adjusting data for <sup>235</sup>U, <sup>238</sup>U, <sup>239</sup>Pu and <sup>240</sup>Pu in conjunction with a class of integral parameters (effective multiplication factors) of fast criticalities including the Godiva and Jezebel spheres, for which the effect of varying resonance parameters in the calculations may be largely neglected.</div><div>The computed mean <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span> of 36 values in total, among which seven refer to the assimilation database, is halved, and the variance of the individual <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span>-values decreases by a factor of 6 with respect to the prior values. More explicitly, the posterior values of respectively 1.10 and 3.77 for <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span> and the variance of the individual <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span>-values are much smaller than the corresponding prior values of 2.32 and 22.38, which is significant.</div><div>As regards the adjustment, for <sup>235</sup>U, it is found that <span><math><mrow><mover><mrow><mi>μ</mi></mrow><mrow><mo>¯</mo></mrow></mover></mrow></math></span> and the inelastic scattering cross-section respectively decrease by maximum 3% and 2.6%; <span><math><mrow><mover><mrow><mi>ν</mi></mrow><mrow><mo>¯</mo></mrow></mover></mrow></math></span> is uniformly decreased by maximum 0.053% and the fission spectrum is somewhat harder. For <sup>238</sup>U, <span><math><mrow><mover><mrow><mi>μ</mi></mrow><mrow><mo>¯</mo></mrow></mover></mrow></math></span> increases by maximum 5.4% and the inelastic scattering cross-section is lowered by 1%. For <sup>239</sup>Pu, <span><math><mrow><mover><mrow><mi>ν</mi></mrow><mrow><mo>¯</mo></mrow></mover></mrow></math></span> is uniformly increased by maximum 0.027% and the fission spectrum is slightly softer. For <sup>240</sup>Pu, the fission cross-section is decreased by maximum 3.9%. The proposed adjustments are well within the uncertainty limit of one standard deviation for the
{"title":"Adjusting JEFF-3.3 actinide data using a new, dedicated methodology for selecting suited assimilation database parameters","authors":"Sandro Pelloni, Dimitri Rochman","doi":"10.1016/j.anucene.2025.111246","DOIUrl":"10.1016/j.anucene.2025.111246","url":null,"abstract":"<div><div>A general method is proposed to choose, among several benchmarks, suitable assimilation database parameters to be used in adjustments. All these benchmarks are predetermined keeping in mind specific goals of the data assimilation. The main principle of the novel methodology is that of excluding groups of strongly cross-correlated computed parameters by estimating their posterior values of <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span>. The assimilation tries minimizing the mean <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span> for the entirety of the considered benchmarks, including those target parameters which are not part of the selected assimilation database.</div><div>The methodology currently applied to the JEFF-3.3 library, is demonstrated by adjusting data for <sup>235</sup>U, <sup>238</sup>U, <sup>239</sup>Pu and <sup>240</sup>Pu in conjunction with a class of integral parameters (effective multiplication factors) of fast criticalities including the Godiva and Jezebel spheres, for which the effect of varying resonance parameters in the calculations may be largely neglected.</div><div>The computed mean <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span> of 36 values in total, among which seven refer to the assimilation database, is halved, and the variance of the individual <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span>-values decreases by a factor of 6 with respect to the prior values. More explicitly, the posterior values of respectively 1.10 and 3.77 for <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span> and the variance of the individual <span><math><mrow><msup><mrow><mi>χ</mi></mrow><mn>2</mn></msup></mrow></math></span>-values are much smaller than the corresponding prior values of 2.32 and 22.38, which is significant.</div><div>As regards the adjustment, for <sup>235</sup>U, it is found that <span><math><mrow><mover><mrow><mi>μ</mi></mrow><mrow><mo>¯</mo></mrow></mover></mrow></math></span> and the inelastic scattering cross-section respectively decrease by maximum 3% and 2.6%; <span><math><mrow><mover><mrow><mi>ν</mi></mrow><mrow><mo>¯</mo></mrow></mover></mrow></math></span> is uniformly decreased by maximum 0.053% and the fission spectrum is somewhat harder. For <sup>238</sup>U, <span><math><mrow><mover><mrow><mi>μ</mi></mrow><mrow><mo>¯</mo></mrow></mover></mrow></math></span> increases by maximum 5.4% and the inelastic scattering cross-section is lowered by 1%. For <sup>239</sup>Pu, <span><math><mrow><mover><mrow><mi>ν</mi></mrow><mrow><mo>¯</mo></mrow></mover></mrow></math></span> is uniformly increased by maximum 0.027% and the fission spectrum is slightly softer. For <sup>240</sup>Pu, the fission cross-section is decreased by maximum 3.9%. The proposed adjustments are well within the uncertainty limit of one standard deviation for the","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111246"},"PeriodicalIF":1.9,"publicationDate":"2025-02-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143348271","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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