<div><div>The maritime sector heavily relies on heavy fuel oil as its primary energy source for ship propulsion, which has a significant adverse impact on the environment, underscoring the urgent need for clean, eco-friendly propulsion alternatives. This paper aims to develop an in-depth thermoeconomic modeling of a real-world nuclear marine propulsion system, employing the nuclear ship Savannah as the baseline benchmark. Given the limited availability of data, this study utilizes innovative ideas to precisely characterize the thermodynamic properties and energy flows within the nuclear propulsion, enabling a comprehensive exergoeconomic performance assessment. Upon validating the developed nuclear ship Savannah propulsion model, an extensive analysis is undertaken from the energy, exergy, and exergoeconomic viewpoints, combining the principles of the first and second laws of thermodynamics with the specific exergy costing technique. The nuclear propulsion model is subsequently integrated with both local sensitivity analysis and global sensitivity analysis to examine how output variables respond to variations in different design parameters. Four key thermoeconomic performance indexes of nuclear propulsion including the energy efficiency, exergy efficiency, propulsion power, and total product exergy cost rate are considered as system output variables. The study employs a one-at-a-time approach for local sensitivity analysis and utilizes the variance-based Sobol method for global sensitivity analysis. Within the local sensitivity analysis framework, a novel indicator, termed “dispersion sensitivity index” is introduced to precisely quantify the overall sensitivity of outputs to inputs. This is subsequently compared with the total sensitivity index obtained from the global sensitivity analysis. The energy analysis demonstrates that the high-pressure and low-pressure steam turbines achieve mechanical power outputs of 6.881 MW and 8.209 MW, respectively, with the overall nuclear propulsion efficiency determined to be 26.18 %. The high-pressure steam generator is identified as the primary source of exergy destruction, with a value of 7161.03 kW, while the condensers exhibit the lowest exergy efficiency, around 30.68 %. Additionally, the exergoeconomic evaluation highlights that the high-pressure steam generator bears the highest exergy costs for both fuel and product, at $1490.20 and $1600.70 per hour, respectively, and the highest total operational cost of $110.60 per hour. The sensitivity analysis reveals that the steam flow rate at the high-pressure turbine inlet exerts the greatest influence on energy efficiency, exergy efficiency, and propulsion power with total sensitivity index of 28.5 %, 42.2 %, and 50.34 %, respectively. Conversely, the heat transfer surface area of high-pressure steam generator has the most significant effect on total product exergy cost rate, with a substantial total sensitivity index of 59.34 %. The integration of sensitivi
{"title":"Development of numerical code for an in-depth energy, exergy, exergoeconomic (3-E) assessments, and sensitivity analysis of NS Savannah marine propulsion: A pre-optimization-focused approach","authors":"Navid Delgarm, Mahmoud Rostami Varnousfaaderani, Hamid Farrokhfal, Sajad Ardeshiri","doi":"10.1016/j.nucengdes.2024.113766","DOIUrl":"10.1016/j.nucengdes.2024.113766","url":null,"abstract":"<div><div>The maritime sector heavily relies on heavy fuel oil as its primary energy source for ship propulsion, which has a significant adverse impact on the environment, underscoring the urgent need for clean, eco-friendly propulsion alternatives. This paper aims to develop an in-depth thermoeconomic modeling of a real-world nuclear marine propulsion system, employing the nuclear ship Savannah as the baseline benchmark. Given the limited availability of data, this study utilizes innovative ideas to precisely characterize the thermodynamic properties and energy flows within the nuclear propulsion, enabling a comprehensive exergoeconomic performance assessment. Upon validating the developed nuclear ship Savannah propulsion model, an extensive analysis is undertaken from the energy, exergy, and exergoeconomic viewpoints, combining the principles of the first and second laws of thermodynamics with the specific exergy costing technique. The nuclear propulsion model is subsequently integrated with both local sensitivity analysis and global sensitivity analysis to examine how output variables respond to variations in different design parameters. Four key thermoeconomic performance indexes of nuclear propulsion including the energy efficiency, exergy efficiency, propulsion power, and total product exergy cost rate are considered as system output variables. The study employs a one-at-a-time approach for local sensitivity analysis and utilizes the variance-based Sobol method for global sensitivity analysis. Within the local sensitivity analysis framework, a novel indicator, termed “dispersion sensitivity index” is introduced to precisely quantify the overall sensitivity of outputs to inputs. This is subsequently compared with the total sensitivity index obtained from the global sensitivity analysis. The energy analysis demonstrates that the high-pressure and low-pressure steam turbines achieve mechanical power outputs of 6.881 MW and 8.209 MW, respectively, with the overall nuclear propulsion efficiency determined to be 26.18 %. The high-pressure steam generator is identified as the primary source of exergy destruction, with a value of 7161.03 kW, while the condensers exhibit the lowest exergy efficiency, around 30.68 %. Additionally, the exergoeconomic evaluation highlights that the high-pressure steam generator bears the highest exergy costs for both fuel and product, at $1490.20 and $1600.70 per hour, respectively, and the highest total operational cost of $110.60 per hour. The sensitivity analysis reveals that the steam flow rate at the high-pressure turbine inlet exerts the greatest influence on energy efficiency, exergy efficiency, and propulsion power with total sensitivity index of 28.5 %, 42.2 %, and 50.34 %, respectively. Conversely, the heat transfer surface area of high-pressure steam generator has the most significant effect on total product exergy cost rate, with a substantial total sensitivity index of 59.34 %. The integration of sensitivi","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113766"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167561","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113796
B. Rossaert , C. Bojanowski , A. Leenaers , G. Cornelis , E. Feldman , E. Wilson , S.Van Dyck , J. Stevens , J. Wight
Experimental research reactor fuel testing is conducted in the Belgian Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) in dedicated irradiation vehicles or rigs. One such vehicle allows flat full-size fuel plates to be irradiated by inserting them into slotted baskets that captures a narrow portion of the longitudinal edges of the plates. The motion of the fuel plates within the baskets is possible within the narrow slots and thus, the plate is considered to be unattached. The design intentionally omits fixing mechanisms of the fuel plates to the baskets to facilitate the inspection and repositioning of the plates between the irradiation cycles and the accommodation of thermal expansion of the plates in the lateral direction. However, loosely inserted fuel plates have weak structural boundary conditions allowing for larger out-of-plane deflections caused by hydrodynamic loads exerted by the flowing coolant, as compared to those of fixed plates. Unexpected large deformations of plates occurred in several irradiation cycles that further resulted in a loss of cladding integrity. These deformations could not be attributed to a single source. This triggered a series of thermal hydraulic, structural, and fluid–structure interaction analyses aiming at understanding the observed phenomenon. The analyses revealed that, for a certain combination of unfavorable manufacturing and assembly tolerances, fuel plate edges could escape out of the slots in the irradiation basket due to the hydrodynamic load. Subsequently, the plate could become wedged inside the basket coolant channel opening. This resulted in reduced coolant flow and accelerated temperature increase and thermal expansion of the plate while under irradiation. This unfavorable feedback loop could then lead to excessive plate surface temperatures, deformed plates and cladding failure, as was observed in the experiments. These analyses not only provided a probable cause of the fuel plate failures, but also resulted in a new and improved design of the irradiation basket to avoid these issues in the future. A series of recent successful irradiations confirm that the sources of failures were identified correctly, and the implemented mitigations were adequate.
{"title":"Failure investigation and mitigation after experimental research reactor fuel plate deformation in an irradiation device","authors":"B. Rossaert , C. Bojanowski , A. Leenaers , G. Cornelis , E. Feldman , E. Wilson , S.Van Dyck , J. Stevens , J. Wight","doi":"10.1016/j.nucengdes.2024.113796","DOIUrl":"10.1016/j.nucengdes.2024.113796","url":null,"abstract":"<div><div>Experimental research reactor fuel testing is conducted in the Belgian Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) in dedicated irradiation vehicles or rigs. One such vehicle allows flat full-size fuel plates to be irradiated by inserting them into slotted baskets that captures a narrow portion of the longitudinal edges of the plates. The motion of the fuel plates within the baskets is possible within the narrow slots and thus, the plate is considered to be unattached. The design intentionally omits fixing mechanisms of the fuel plates to the baskets to facilitate the inspection and repositioning of the plates between the irradiation cycles and the accommodation of thermal expansion of the plates in the lateral direction. However, loosely inserted fuel plates have weak structural boundary conditions allowing for larger out-of-plane deflections caused by hydrodynamic loads exerted by the flowing coolant, as compared to those of fixed plates. Unexpected large deformations of plates occurred in several irradiation cycles that further resulted in a loss of cladding integrity. These deformations could not be attributed to a single source. This triggered a series of thermal hydraulic, structural, and fluid–structure interaction analyses aiming at understanding the observed phenomenon. The analyses revealed that, for a certain combination of unfavorable manufacturing and assembly tolerances, fuel plate edges could escape out of the slots in the irradiation basket due to the hydrodynamic load. Subsequently, the plate could become wedged inside the basket coolant channel opening. This resulted in reduced coolant flow and accelerated temperature increase and thermal expansion of the plate while under irradiation. This unfavorable feedback loop could then lead to excessive plate surface temperatures, deformed plates and cladding failure, as was observed in the experiments. These analyses not only provided a probable cause of the fuel plate failures, but also resulted in a new and improved design of the irradiation basket to avoid these issues in the future. A series of recent successful irradiations confirm that the sources of failures were identified correctly, and the implemented mitigations were adequate.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113796"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168326","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113749
Binzhuo Xia, Kui Zhang, Ronghua Chen, Wenxi Tian, Suizheng Qiu
Because of exceptional economic benefits, stability, and high energy production efficiency, marine nuclear reactors are a major technology reserve for marine power supply. The Marine Nuclear Reactor’s operating safety will be put to the test by the motion conditions it encounters. It is essential to investigate the critical heat flux characteristic under motion conditions since it is a crucial thermal hydraulic feature of nuclear reactor cores and one of the key criteria for thermal hydraulic design. Some CHF mechanisms under motion conditions have been obtained, and the CHF features of marine nuclear reactor cores under motion conditions have become a number of study issues, including model development and CHF experiments. Still, a few significant problems need to be addressed. In light of this, the experimental investigation and the development of the CHF prediction methods under motion conditions are summarized and analyzed in this work. This paper reviews the present correlations of CHF and the prediction methods created based on mechanism models. It also summarizes the experimental progress of the impacts of inclined motion, rolling motion, heaving motion, and coupling motion on CHF. Ultimately, this research field’s development prospects are outlined and projected.
{"title":"Review of CHF experiment and prediction methods under motion condition","authors":"Binzhuo Xia, Kui Zhang, Ronghua Chen, Wenxi Tian, Suizheng Qiu","doi":"10.1016/j.nucengdes.2024.113749","DOIUrl":"10.1016/j.nucengdes.2024.113749","url":null,"abstract":"<div><div>Because of exceptional economic benefits, stability, and high energy production efficiency, marine nuclear reactors are a major technology reserve for marine power supply. The Marine Nuclear Reactor’s operating safety will be put to the test by the motion conditions it encounters. It is essential to investigate the critical heat flux characteristic under motion conditions since it is a crucial thermal hydraulic feature of nuclear reactor cores and one of the key criteria for thermal hydraulic design. Some CHF mechanisms under motion conditions have been obtained, and the CHF features of marine nuclear reactor cores under motion conditions have become a number of study issues, including model development and CHF experiments. Still, a few significant problems need to be addressed. In light of this, the experimental investigation and the development of the CHF prediction methods under motion conditions are summarized and analyzed in this work. This paper reviews the present correlations of CHF and the prediction methods created based on mechanism models. It also summarizes the experimental progress of the impacts of inclined motion, rolling motion, heaving motion, and coupling motion on CHF. Ultimately, this research field’s development prospects are outlined and projected.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113749"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168328","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113813
G. Vikram , Amit Kumar Chauhan , M. Rajendrakumar , K. Natesan
The primary circuit of a typical pool-type sodium-cooled fast reactor (SFR) is a complex flow network with multiple flow paths. The resistances of these flow paths vary significantly from very low to very high values. Estimation of flow fractions in these paths is essential for the design and analysis of various components in the primary circuit under various operating conditions. In the present work, the primary circuit of a typical medium-sized pool-type SFR has been modeled using the Flownex code, a commercial system dynamics code. The steady-state flow distribution in the primary circuit has been first studied and an overall flow balance has been established. Out of the total flow supplied by the primary pumps, ∼91 % flows through the core, and ∼ 93 % flows through the Intermediate Heat Exchanger (IHX). Notably, in the storage locations of the core, there is no leakage from the Grid Plate (GP, the structure on which the core subassemblies (SAs) are supported) to the bottom Core Support Structure (CSS) plenum. Instead, flow is in the reverse direction due to the high resistance offered by the sleeve holes in the storage locations.
Then, the effect of removing SAs from the fuel and storage locations of the core has been studied. The empty sleeves in the GP were modeled in 3D using Ansys® Fluent and have been coupled with the Flownex model of the primary circuit. Two different flow configurations were observed in the empty GP sleeves when the fuel SAs were removed and when the storage SAs were removed. When fuel SAs are removed, the sodium flows downwards in the empty sleeve bottom opening. However, when storage SAs are removed, the sodium flows upwards in the empty sleeve bottom opening. This happens because of the high hydraulic resistance offered by the storage SA sleeve holes. As a result, when fuel SAs are removed, the flow rates in the paths fed by the CSS plenum (main vessel cooling system path, shielding SA flow path, etc.) increase. When storage SAs are removed, the flows in these paths decrease. There is no significant change in the pump operating point when a single fuel or storage SA is removed from the core. However, when more fuel SAs are removed from the core, a change in the pump operating point is observed. For instance, the pump flow decreases by ∼ 4.1 % when seven fuel SAs are removed from the core.
{"title":"Flow distribution in the primary circuit of a fast reactor: Impact with reduced number of subassemblies in the core","authors":"G. Vikram , Amit Kumar Chauhan , M. Rajendrakumar , K. Natesan","doi":"10.1016/j.nucengdes.2024.113813","DOIUrl":"10.1016/j.nucengdes.2024.113813","url":null,"abstract":"<div><div>The primary circuit of a typical pool-type sodium-cooled fast reactor (SFR) is a complex flow network with multiple flow paths. The resistances of these flow paths vary significantly from very low to very high values. Estimation of flow fractions in these paths is essential for the design and analysis of various components in the primary circuit under various operating conditions. In the present work, the primary circuit of a typical medium-sized pool-type SFR has been modeled using the Flownex code, a commercial system dynamics code. The steady-state flow distribution in the primary circuit has been first studied and an overall flow balance has been established. Out of the total flow supplied by the primary pumps, ∼91 % flows through the core, and ∼ 93 % flows through the Intermediate Heat Exchanger (IHX). Notably, in the storage locations of the core, there is no leakage from the Grid Plate (GP, the structure on which the core subassemblies (SAs) are supported) to the bottom Core Support Structure (CSS) plenum. Instead, flow is in the reverse direction due to the high resistance offered by the sleeve holes in the storage locations.</div><div>Then, the effect of removing SAs from the fuel and storage locations of the core has been studied. The empty sleeves in the GP were modeled in 3D using Ansys<em>®</em> Fluent and have been coupled with the Flownex model of the primary circuit. Two different flow configurations were observed in the empty GP sleeves when the fuel SAs were removed and when the storage SAs were removed. When fuel SAs are removed, the sodium flows downwards in the empty sleeve bottom opening. However, when storage SAs are removed, the sodium flows upwards in the empty sleeve bottom opening. This happens because of the high hydraulic resistance offered by the storage SA sleeve holes. As a result, when fuel SAs are removed, the flow rates in the paths fed by the CSS plenum (main vessel cooling system path, shielding SA flow path, etc.) increase. When storage SAs are removed, the flows in these paths decrease. There is no significant change in the pump operating point when a single fuel or storage SA is removed from the core. However, when more fuel SAs are removed from the core, a change in the pump operating point is observed. For instance, the pump flow decreases by ∼ 4.1 % when seven fuel SAs are removed from the core.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113813"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168334","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113788
Tianchi Li , Zengliang Mo , Jia Zhou , Qi Chen , Zhi Cao , Jianhua Guo , Zhongyuan Yang , Chunwei Tang , Wensi Li , Yuzhou Ming , Fang Liu , Taihong Yan , Gaoyang Mi , Weifang Zheng
Utilizing laser technology for the dismantling and cutting of fast reactor assembly represents a viable nuclear fuel reprocessing technology for future applications. The commonly adopted procedure involves removing the hexagonal tube and end structures of the assembly without compromising the integrity of the fuel rods, which are then cut into short segments. Therefore, mastering the laser cutting parameters that ensure high-quality cuts of the hexagonal tube while minimizing damage to the internal component rods is essential. This study investigates the impact of laser cutting parameters on the quality of cuts in stainless steel hexagonal tubes. Optimal conditions—3.5 m/min cutting speed, −1.5 mm focal position, 4800 W power, and nitrogen at 15 MPa—produced minimal kerf width (0.438 mm), surface roughness (4.21 μm), and slagging length (0.206 mm). These findings highlight the importance of precise parameter control in laser cutting for nuclear applications, offering significant improvements in efficiency and safety for fast reactor fuel reprocessing.
{"title":"Experimental study on the laser cutting process of the stainless steel hexagonal tube of fast reactor simulate assembly","authors":"Tianchi Li , Zengliang Mo , Jia Zhou , Qi Chen , Zhi Cao , Jianhua Guo , Zhongyuan Yang , Chunwei Tang , Wensi Li , Yuzhou Ming , Fang Liu , Taihong Yan , Gaoyang Mi , Weifang Zheng","doi":"10.1016/j.nucengdes.2024.113788","DOIUrl":"10.1016/j.nucengdes.2024.113788","url":null,"abstract":"<div><div>Utilizing laser technology for the dismantling and cutting of fast reactor assembly represents a viable nuclear fuel reprocessing technology for future applications. The commonly adopted procedure involves removing the hexagonal tube and end structures of the assembly without compromising the integrity of the fuel rods, which are then cut into short segments. Therefore, mastering the laser cutting parameters that ensure high-quality cuts of the hexagonal tube while minimizing damage to the internal component rods is essential. This study investigates the impact of laser cutting parameters on the quality of cuts in stainless steel hexagonal tubes. Optimal conditions—3.5 m/min cutting speed, −1.5 mm focal position, 4800 W power, and nitrogen at 15 MPa—produced minimal kerf width (0.438 mm), surface roughness (4.21 μm), and slagging length (0.206 mm). These findings highlight the importance of precise parameter control in laser cutting for nuclear applications, offering significant improvements in efficiency and safety for fast reactor fuel reprocessing.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113788"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168374","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113798
Yantao Luo , Tian Zhang , Antonio Cammi , Xiang Wang
Common pool-type heating reactor designs have certain safety risks, particularly during natural disasters such as earthquakes. In such scenarios, if the pool water leaks, the reactor core is likely exposed to air, which can cause core meltdown. The proposed expandable LOTUS reactor offers a solution to this problem. Its core can unfold under the influence of gravity and buoyancy, passively shut down the reactor and remove the core heat when accidents occur. During the unfolding process, both geometric buckling and material buckling of the core change non-linearly, and this is a new problem for nuclear engineering that was not discussed yet. Therefore, the emergence of the LOTUS reactor entails an entirely new and intricate nuclear-thermal transient coupling process through numerical simulation. We used SERPENT and STAR-CCM + to conduct single-physical-field and nuclear-thermal coupling calculations. The former provides a fast, low-dimensional calculation that serves as references for the coupling calculation. The changes in power, effective multiplication factor and temperature indicate that the unfolding motion can force the reactor to shut down, and the remaining pool water effectively decreases the core temperature. For the coupling calculations, first, validation calculations confirmed the correctness and stability of the coupling strategy. Then, steady-state coupling calculations were conducted using Picard iteration method and converged after 207 iterations. There was a maximal power difference of up to 6 % between single-physical-field calculations and coupling calculations at the beginning of the unfolding motion. This difference gradually decreased over time and disappeared when the unfolding angle reached 0.6°. Compared to the single-physical calculations, the non-uniformly distributed temperature obtained from the coupling calculations was more consistent with the actual distribution. These differences highlight the necessity and importance of coupling calculations.
{"title":"Neutronic and thermal-hydraulic coupling analysis of the dynamic unfolding process of the LOTUS reactor","authors":"Yantao Luo , Tian Zhang , Antonio Cammi , Xiang Wang","doi":"10.1016/j.nucengdes.2024.113798","DOIUrl":"10.1016/j.nucengdes.2024.113798","url":null,"abstract":"<div><div>Common pool-type heating reactor designs have certain safety risks, particularly during natural disasters such as earthquakes. In such scenarios, if the pool water leaks, the reactor core is likely exposed to air, which can cause core meltdown. The proposed expandable LOTUS reactor offers a solution to this problem. Its core can unfold under the influence of gravity and buoyancy, passively shut down the reactor and remove the core heat when accidents occur. During the unfolding process, both geometric buckling and material buckling of the core change non-linearly, and this is a new problem for nuclear engineering that was not discussed yet. Therefore, the emergence of the LOTUS reactor entails an entirely new and intricate nuclear-thermal transient coupling process through numerical simulation. We used SERPENT and STAR-CCM + to conduct single-physical-field and nuclear-thermal coupling calculations. The former provides a fast, low-dimensional calculation that serves as references for the coupling calculation. The changes in power, effective multiplication factor and temperature indicate that the unfolding motion can force the reactor to shut down, and the remaining pool water effectively decreases the core temperature. For the coupling calculations, first, validation calculations confirmed the correctness and stability of the coupling strategy. Then, steady-state coupling calculations were conducted using Picard iteration method and converged after 207 iterations. There was a maximal power difference of up to 6 % between single-physical-field calculations and coupling calculations at the beginning of the unfolding motion. This difference gradually decreased over time and disappeared when the unfolding angle reached 0.6°. Compared to the single-physical calculations, the non-uniformly distributed temperature obtained from the coupling calculations was more consistent with the actual distribution. These differences highlight the necessity and importance of coupling calculations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113798"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168376","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113811
J.A. Hirschhorn , L.K. Aagesen , C. Jiang , G.L. Beausoleil II
Despite decades of fuel rod material and design improvements, fuel-cladding chemical interaction (FCCI) remains the single-most lifetime-limiting behavior for modern metallic fuel rods. Constraining fuel lifetime increases operating costs, limiting the economic viability of commercializing metallic nuclear fuel technology. A mechanistic multiscale model utilizing the finite element method-based MARMOT and BISON codes was developed to more confidently predict cladding-side FCCI and its impact on fuel performance. The new BISON model incorporates mesoscale models for the effects of fuel microstructure evolution on the transport of wastage-inducing lanthanides through the fuel and for the kinetics of cladding wastage layer growth. The mesoscale models, in turn, build on lanthanide transport property data obtained from the atomistic scale. Preliminary validation studies using wastage thickness and cladding profilometry data from four fuel rods irradiated in Experimental Breeder Reactor II experiment X447 and one fuel rod from Fast Flux Test Facility experiment IFR1 show that the new model predicts cladding wastage and its effects on cladding deformation as well as existing empirical FCCI correlations. The new model is expected to aid in the design of new metallic fuel concepts, including fuel additives, cladding liners, and sodium-free annular fuel geometries. Future work will focus on broader validation and refinement of the model’s treatment of different fuel alloys and cladding materials.
{"title":"Development and preliminary validation of a mechanistic multiscale model for fuel-cladding chemical interaction in metallic nuclear fuels","authors":"J.A. Hirschhorn , L.K. Aagesen , C. Jiang , G.L. Beausoleil II","doi":"10.1016/j.nucengdes.2024.113811","DOIUrl":"10.1016/j.nucengdes.2024.113811","url":null,"abstract":"<div><div>Despite decades of fuel rod material and design improvements, fuel-cladding chemical interaction (FCCI) remains the single-most lifetime-limiting behavior for modern metallic fuel rods. Constraining fuel lifetime increases operating costs, limiting the economic viability of commercializing metallic nuclear fuel technology. A mechanistic multiscale model utilizing the finite element method-based MARMOT and BISON codes was developed to more confidently predict cladding-side FCCI and its impact on fuel performance. The new BISON model incorporates mesoscale models for the effects of fuel microstructure evolution on the transport of wastage-inducing lanthanides through the fuel and for the kinetics of cladding wastage layer growth. The mesoscale models, in turn, build on lanthanide transport property data obtained from the atomistic scale. Preliminary validation studies using wastage thickness and cladding profilometry data from four fuel rods irradiated in Experimental Breeder Reactor II experiment X447 and one fuel rod from Fast Flux Test Facility experiment IFR1 show that the new model predicts cladding wastage and its effects on cladding deformation as well as existing empirical FCCI correlations. The new model is expected to aid in the design of new metallic fuel concepts, including fuel additives, cladding liners, and sodium-free annular fuel geometries. Future work will focus on broader validation and refinement of the model’s treatment of different fuel alloys and cladding materials.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113811"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169075","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113815
Irena Kratochvílová , Lucie Celbová , Přemysl Vaněk , Dagmar Chvostová , Ubaid Ahmed , Daniel Šimek , Stanislav Cichoň , Tomáš Peltan , Stevan Gavranović , Maksym Buryi , David John , Radek Škoda , Jan Pospíšil
Energy harvesting is a versatile approach that holds promise for generating clean energy and enhancing the sustainability of infrastructure. Given the challenges associated with nuclear waste, such as radiation levels and long-term storage requirements, developing suitable materials for energy harvesting is crucial. These materials must be capable of withstanding the harsh conditions present in nuclear waste storage facilities while efficiently capturing and converting energy. The objective of this work was to study lead halide perovskites (MAPbCl3 and MAPbI3.) in terms of their application for conversion and storage of energy released from spent nuclear fuel. Firstly, in this work, the detailed analysis of MAPbCl3 and MAPbI3 irradiated by spent nuclear fuel assembly IRT–4 M (consisting of high energy gamma photons and suppressed neutron fluxes) was done. The spent nuclear fuel irradiation practically did not change the materials structures, composition, chemical bonds, phase transitions (temperatures, energies). On the other side, irradiation caused the defects in the electron shells. Our results confirm the possibility of using the investigated perovskites for the conversion and storage of energy from spent nuclear fuel: 1. to convert the energy of spent nuclear fuel assembly into release of the electrical charge (photogenerated electron-hole pairs) and 2. to store the spent fuel energy into perovskite high temperature crystal phase. When the radiation increases the temperature up to high temperature first order phase transition the energy is consumed/stored into the high temperature material phase. As wide range of perovskite materials have very diverse first order phase transition temperatures it is possible to match the specific spent nuclear fuel radiation dose with the proper perovskite phase transition temperature. Finally, the photogenerated charges should efficiently extracted from the material and the stored energy can be released and subsequently reused by lowering the temperature (typically after removing the perovskite from the spent nuclear fuel cask) until the material returns to the low-temperature phase.
{"title":"Perovskites as potential candidates for storage and conversion of spent nuclear fuel energy","authors":"Irena Kratochvílová , Lucie Celbová , Přemysl Vaněk , Dagmar Chvostová , Ubaid Ahmed , Daniel Šimek , Stanislav Cichoň , Tomáš Peltan , Stevan Gavranović , Maksym Buryi , David John , Radek Škoda , Jan Pospíšil","doi":"10.1016/j.nucengdes.2024.113815","DOIUrl":"10.1016/j.nucengdes.2024.113815","url":null,"abstract":"<div><div>Energy harvesting is a versatile approach that holds promise for generating clean energy and enhancing the sustainability of infrastructure. Given the challenges associated with nuclear waste, such as radiation levels and long-term storage requirements, developing suitable materials for energy harvesting is crucial. These materials must be capable of withstanding the harsh conditions present in nuclear waste storage facilities while efficiently capturing and converting energy. The objective of this work was to study lead halide perovskites (MAPbCl<sub>3</sub> and MAPbI<sub>3</sub>.) in terms of their application for conversion and storage of energy released from spent nuclear fuel. Firstly, in this work, the detailed analysis of MAPbCl<sub>3</sub> and MAPbI<sub>3</sub> irradiated by spent nuclear fuel assembly IRT–4 M (consisting of high energy gamma photons and suppressed neutron fluxes) was done. The spent nuclear fuel irradiation practically did not change the materials structures, composition, chemical bonds, phase transitions (temperatures, energies). On the other side, irradiation caused the defects in the electron shells. Our results confirm the possibility of using the investigated perovskites for the conversion and storage of energy from spent nuclear fuel: 1. to convert the energy of spent nuclear fuel assembly into release of the electrical charge (photogenerated electron-hole pairs) and 2. to store the spent fuel energy into perovskite high temperature crystal phase. When the radiation increases the temperature up to high temperature first order phase transition the energy is consumed/stored into the high temperature material phase. As wide range of perovskite materials have very diverse first order phase transition temperatures it is possible to match the specific spent nuclear fuel radiation dose with the proper perovskite phase transition temperature. Finally, the photogenerated charges should efficiently extracted from the material and the stored energy can be released and subsequently reused by lowering the temperature (typically after removing the perovskite from the spent nuclear fuel cask) until the material returns to the low-temperature phase.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113815"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169079","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113729
Fatima Ghandour , Ziad Francis
CAREM 25 is considered one of the early versions of Small Modular Reactors (SMRs) worldwide. Despite efforts to simulate CAREM using black box Monte-Carlo codes, this paper is the first to simulate CAREM 25 or any other core reactor with full complex geometries using open-source software, such as GEANT4. Recent studies have demonstrated that GEANT4 can effectively analyze neutron transport in nuclear physics and engineering. This study aimed to illustrate the potential of GEANT4 in evaluating power peaking factors (PPFs) in the CAREM reactor compared to previous studies. We calculated the power peaking factors (PPFs) of the fuel assemblies and fuel rods using the UserAction classes and the built-in ROOT toolkit in GEANT4. The results showed a flattening of the power distribution, which ensures less disparity between regions with the highest power density (the central fuel assembly with a PPF of 1.502) and other fuel assemblies. This promotes stable reactor behavior while maintaining safety margins. We calculated the detailed PPFs of the 108 fuel rods in both the hottest assembly and the average assembly (with a PPF of 1.005), revealing a flattened power distribution across the fuel assemblies. While direct validation against MCNPX was not achievable due to variations in conditions, our investigation identified consistent trends in the PPFs of both GEANT4 and MCNPX, suggesting similar behavior across the fuel assemblies. This research takes an important step in employing open-source codes like GEANT4 in reactor physics with complex core geometries, with further investigations needed to enhance its effectiveness and alignment with other Monte Carlo codes.
{"title":"A novel approach to estimate the power peaking factors in a complex geometry iPWR inspired by CAREM 25 using GEANT4 toolkit","authors":"Fatima Ghandour , Ziad Francis","doi":"10.1016/j.nucengdes.2024.113729","DOIUrl":"10.1016/j.nucengdes.2024.113729","url":null,"abstract":"<div><div>CAREM 25 is considered one of the early versions of Small Modular Reactors (SMRs) worldwide. Despite efforts to simulate CAREM using black box Monte-Carlo codes, this paper is the first to simulate CAREM 25 or any other core reactor with full complex geometries using open-source software, such as GEANT4. Recent studies have demonstrated that GEANT4 can effectively analyze neutron transport in nuclear physics and engineering. This study aimed to illustrate the potential of GEANT4 in evaluating power peaking factors (PPFs) in the CAREM reactor compared to previous studies. We calculated the power peaking factors (PPFs) of the fuel assemblies and fuel rods using the UserAction classes and the built-in ROOT toolkit in GEANT4. The results showed a flattening of the power distribution, which ensures less disparity between regions with the highest power density (the central fuel assembly with a PPF of 1.502) and other fuel assemblies. This promotes stable reactor behavior while maintaining safety margins. We calculated the detailed PPFs of the 108 fuel rods in both the hottest assembly and the average assembly (with a PPF of 1.005), revealing a flattened power distribution across the fuel assemblies. While direct validation against MCNPX was not achievable due to variations in conditions, our investigation identified consistent trends in the PPFs of both GEANT4 and MCNPX, suggesting similar behavior across the fuel assemblies. This research takes an important step in employing open-source codes like GEANT4 in reactor physics with complex core geometries, with further investigations needed to enhance its effectiveness and alignment with other Monte Carlo codes.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113729"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169083","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 : 2025-02-01DOI: 10.1016/j.nucengdes.2024.113787
Ramiro Freile , Peter German, Mauricio Tano
This article studies the impact of Reynolds Averaged Navier Stokes (RANS) turbulence modeling for the thermal–hydraulics analysis of open-core molten salt reactors (MSRs). The first part of the article describes the and SST RANS turbulence models recently introduced to Pronghorn in the Idaho National Laboratory’s (INL) Multiphysics Object-Oriented Simulation Environment (MOOSE). Special attention is given to the near-wall flow modeling strategies implemented in both models and their computational implementation in MOOSE. Both models are validated for canonical experimental data that resemble the flow phenomena expected for MSRs: a channel flow, a backward facing step, and swirling flow in a curved pipe. These models are then applied to the open-source specifications of the Molten Chloride Reactor Experiment (MCRE), which is named Lotus Molten Salt Reactor (L-MSR) for conciseness. First, the flow phenomena and pressure drops predicted by both models are compared with higher-fidelity LES simulations for the expected reactor operational conditions. Then, the impact of the selected turbulence model on steady-state temperature fields distribution is evaluated. Finally, recommendations are given for the selected approach to turbulence modeling of open-core MSRs.
{"title":"Evaluating turbulence modeling for thermal–hydraulics analysis of molten salt reactors","authors":"Ramiro Freile , Peter German, Mauricio Tano","doi":"10.1016/j.nucengdes.2024.113787","DOIUrl":"10.1016/j.nucengdes.2024.113787","url":null,"abstract":"<div><div>This article studies the impact of Reynolds Averaged Navier Stokes (RANS) turbulence modeling for the thermal–hydraulics analysis of open-core molten salt reactors (MSRs). The first part of the article describes the <span><math><mrow><mi>k</mi><mo>−</mo><mi>ϵ</mi></mrow></math></span> and <span><math><mrow><mi>k</mi><mo>−</mo><mi>ω</mi></mrow></math></span> SST RANS turbulence models recently introduced to Pronghorn in the Idaho National Laboratory’s (INL) Multiphysics Object-Oriented Simulation Environment (MOOSE). Special attention is given to the near-wall flow modeling strategies implemented in both models and their computational implementation in MOOSE. Both models are validated for canonical experimental data that resemble the flow phenomena expected for MSRs: a channel flow, a backward facing step, and swirling flow in a curved pipe. These models are then applied to the open-source specifications of the Molten Chloride Reactor Experiment (MCRE), which is named Lotus Molten Salt Reactor (L-MSR) for conciseness. First, the flow phenomena and pressure drops predicted by both models are compared with higher-fidelity LES simulations for the expected reactor operational conditions. Then, the impact of the selected turbulence model on steady-state temperature fields distribution is evaluated. Finally, recommendations are given for the selected approach to turbulence modeling of open-core MSRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113787"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167563","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}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：