Pub Date : 2025-02-26DOI: 10.1016/j.anucene.2025.111291
Karytha M.S. Corrêa , Natália G.P.L. Oliveira , Claubia Pereira , Carlos E. Velasquez
This study proposes evaluating hybrid fusion-fission systems based on the ARC reactor, which uses reprocessed fuels in the fission transmutation layer. The analysis compared two reprocessed fuels: one based on dioxide ((TRU,Th)O2) and the other based on nitride ((TRU,Th)N). The analysis focused on the neutronic parameters and the burnup of these fuels. The MCNP particle transport code was used for the simulations, while the burnup process of the fuels in the (TRU,Th)O2 and (TRU,Th)N systems was conducted using the MONTEBURNS code, which links MCNP and the ORIGEN2.1 depletion code. The results indicate that the (TRU,Th)N system may exhibit superior efficiency in the transmutation of minor actinides during fuel burnup compared to the dioxide-based system.
{"title":"Analysis of reprocessed fuels inserted into a hybrid fusion-fission system based on the ARC reactor","authors":"Karytha M.S. Corrêa , Natália G.P.L. Oliveira , Claubia Pereira , Carlos E. Velasquez","doi":"10.1016/j.anucene.2025.111291","DOIUrl":"10.1016/j.anucene.2025.111291","url":null,"abstract":"<div><div>This study proposes evaluating hybrid fusion-fission systems based on the ARC reactor, which uses reprocessed fuels in the fission transmutation layer. The analysis compared two reprocessed fuels: one based on dioxide ((TRU,Th)O<sub>2</sub>) and the other based on nitride ((TRU,Th)N). The analysis focused on the neutronic parameters and the burnup of these fuels. The MCNP particle transport code was used for the simulations, while the burnup process of the fuels in the (TRU,Th)O<sub>2</sub> and (TRU,Th)N systems was conducted using the MONTEBURNS code, which links MCNP and the ORIGEN2.1 depletion code. The results indicate that the (TRU,Th)N system may exhibit superior efficiency in the transmutation of minor actinides during fuel burnup compared to the dioxide-based system.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111291"},"PeriodicalIF":1.9,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143488134","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 : 2025-02-25DOI: 10.1016/j.anucene.2025.111303
Huifang Zhang , Yuntao Zheng , Yanyu Sun , Kui Ge , Shihao Wu , Yongkang Li , Yapei Zhang , Wenxi Tian , Suizheng Qiu , Guanghui Su
In advanced pressurized water reactors (PWR) employing the passive containment air cooling system (PAS), heat can be removed from the containment to the atmosphere. This paper established a comprehensive mathematical and physical model to simulate PAS and developed a modular PAS module based on the ATHROC to analyze its thermal–hydraulic characteristics. The accuracy of the code was validated by comparing the calculated temperatures and pressures with data obtained from GOTHIC. The results showed that differences were less than 5%, confirming the reliability of the PAS module. Then the thermo-hydraulic analysis of AP1000 containment system with PAS during a loss-of-coolant accident (LOCA) was performed using ATHROC. The findings indicated that when LOCA ocurs, the internal temperature and pressure of the containment can be maintained below the design limits. This study confirms that the PAS module possesses the capability of the thermal–hydraulic analysis of the passive containment air cooling system.
{"title":"Development of a PAS module in the ATHROC for simulation of passive containment air cooling system","authors":"Huifang Zhang , Yuntao Zheng , Yanyu Sun , Kui Ge , Shihao Wu , Yongkang Li , Yapei Zhang , Wenxi Tian , Suizheng Qiu , Guanghui Su","doi":"10.1016/j.anucene.2025.111303","DOIUrl":"10.1016/j.anucene.2025.111303","url":null,"abstract":"<div><div>In advanced pressurized water reactors (PWR) employing the passive containment air cooling system (PAS), heat can be removed from the containment to the atmosphere. This paper established a comprehensive mathematical and physical model to simulate PAS and developed a modular PAS module based on the ATHROC to analyze its thermal–hydraulic characteristics. The accuracy of the code was validated by comparing the calculated temperatures and pressures with data obtained from GOTHIC. The results showed that differences were less than 5%, confirming the reliability of the PAS module. Then the thermo-hydraulic analysis of AP1000 containment system with PAS during a loss-of-coolant accident (LOCA) was performed using ATHROC. The findings indicated that when LOCA ocurs, the internal temperature and pressure of the containment can be maintained below the design limits. This study confirms that the PAS module possesses the capability of the thermal–hydraulic analysis of the passive containment air cooling system.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111303"},"PeriodicalIF":1.9,"publicationDate":"2025-02-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143480420","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 : 2025-02-25DOI: 10.1016/j.anucene.2025.111306
Guozhong Zheng , Rongxin Ni , Xinyu Li , Yuan Gao , Sihan Liu
The rapid expansion of the nuclear energy industry is accompanied by a corresponding increase in the production of spent nuclear fuel (SNF). Extensive studies were conducted on dry storage systems (DSS) in response to the increasing requirements for SNF storage. The main objective of this study is to provide a comprehensive overview of the recent advancements in the field of CFD simulation methods and thermal hydraulic property for SNF dry storage casks. Firstly, the design features and operational mechanisms of the dry storage cask are briefly introduced. Secondly, the simulation methods employed by different researchers are explored and summarized. Finally, the research progresses on heat flow and temperature distribution characteristics, backfill materials and long term storage changes are discussed. The conclusion demonstrates that dry storage casks provide a safe, practical and cost-effective solution for meeting the evolving storage requirements of SNF.
{"title":"Overview of CFD simulation methods and thermal hydraulic property for spent nuclear fuel dry storage cask","authors":"Guozhong Zheng , Rongxin Ni , Xinyu Li , Yuan Gao , Sihan Liu","doi":"10.1016/j.anucene.2025.111306","DOIUrl":"10.1016/j.anucene.2025.111306","url":null,"abstract":"<div><div>The rapid expansion of the nuclear energy industry is accompanied by a corresponding increase in the production of spent nuclear fuel (SNF). Extensive studies were conducted on dry storage systems (DSS) in response to the increasing requirements for SNF storage. The main objective of this study is to provide a comprehensive overview of the recent advancements in the field of CFD simulation methods and thermal hydraulic property for SNF dry storage casks. Firstly, the design features and operational mechanisms of the dry storage cask are briefly introduced. Secondly, the simulation methods employed by different researchers are explored and summarized. Finally, the research progresses on heat flow and temperature distribution characteristics, backfill materials and long term storage changes are discussed. The conclusion demonstrates that dry storage casks provide a safe, practical and cost-effective solution for meeting the evolving storage requirements of SNF.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111306"},"PeriodicalIF":1.9,"publicationDate":"2025-02-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143480419","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 : 2025-02-25DOI: 10.1016/j.anucene.2025.111295
Junhyoung Gil, Ayman I. Hawari
Triuranium octoxide (U3O8) is a starting material in the nuclear fuel cycle and a preferred form for the safe storage of nuclear fuel and for the disposal and containment of radioactive waste, due to its thermodynamically stable characteristics compared to other uranium oxides. Despite extensive experimental and computational investigations on its properties, discrepancies exist regarding its correct magnetic state. Recent inelastic neutron scattering experiments and density functional theory calculations have elucidated that U3O8 exhibits an antiferromagnetic (AFM) [0.5 1 1] ordering. In this study, we model the α-U3O8 crystal with AFM [0.5 1 1] ordering at the atomic level to evaluate its thermal scattering law (TSL). Ab initio lattice dynamics simulations based on density functional theory are coupled with phonon vibration analysis to derive the phonon density of states of α-U3O8, based on which the scattering cross section is generated. The optimized lattice structure of our α-U3O8 model closely aligns with experimental data, and the electronic band gap reasonably matches the experimental range. Phonon vibration analysis under the harmonic approximation indicates that the heat capacity, entropy, and enthalpy deviate by less than 10% from experimental observations. The calculated phonon density of states (DOS) demonstrates overall reasonable agreement with the data obtained from inelastic neutron scattering measurements. Using the DOS, the TSL and thermal neutron scattering cross sections are produced, and the anticipated systematic behavior is exhibited across various temperatures.
{"title":"Evaluation of the thermal scattering law and cross sections for α-U3O8 using ab initio lattice dynamics","authors":"Junhyoung Gil, Ayman I. Hawari","doi":"10.1016/j.anucene.2025.111295","DOIUrl":"10.1016/j.anucene.2025.111295","url":null,"abstract":"<div><div>Triuranium octoxide (U<sub>3</sub>O<sub>8</sub>) is a starting material in the nuclear fuel cycle and a preferred form for the safe storage of nuclear fuel and for the disposal and containment of radioactive waste, due to its thermodynamically stable characteristics compared to other uranium oxides. Despite extensive experimental and computational investigations on its properties, discrepancies exist regarding its correct magnetic state. Recent inelastic neutron scattering experiments and density functional theory calculations have elucidated that U<sub>3</sub>O<sub>8</sub> exhibits an antiferromagnetic (AFM) [0.5 1 1] ordering. In this study, we model the α-U<sub>3</sub>O<sub>8</sub> crystal with AFM [0.5 1 1] ordering at the atomic level to evaluate its thermal scattering law (TSL). Ab initio lattice dynamics simulations based on density functional theory are coupled with phonon vibration analysis to derive the phonon density of states of α-U<sub>3</sub>O<sub>8</sub>, based on which the scattering cross section is generated. The optimized lattice structure of our α-U<sub>3</sub>O<sub>8</sub> model closely aligns with experimental data, and the electronic band gap reasonably matches the experimental range. Phonon vibration analysis under the harmonic approximation indicates that the heat capacity, entropy, and enthalpy deviate by less than 10% from experimental observations. The calculated phonon density of states (DOS) demonstrates overall reasonable agreement with the data obtained from inelastic neutron scattering measurements. Using the DOS, the TSL and thermal neutron scattering cross sections are produced, and the anticipated systematic behavior is exhibited across various temperatures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111295"},"PeriodicalIF":1.9,"publicationDate":"2025-02-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143480440","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 : 2025-02-24DOI: 10.1016/j.anucene.2025.111270
Lu Li, Guoqiang Wei, Junnan Zhang, Jingyun Huang, Qingning Zuo, Shuhan Zhuang, Xiaoping Bai
The paper takes radionuclide in the reactor building of PWR under normal condition as the research object, a set of nuclide balance equations is established through analyzing the production and disappearance terms of radionuclide in the reactor building. According to the intermittent operation of small sweep gas and introduction of the decay chain, both periodic iteration method and transmutation trajectory analysis method (TTA) are employed to solve for the quantity of radioactive nuclides in the reactor building. Building a virtual nuclear power plant as an example, the radioactive source terms released to the environment from the reactor building is analyzed based on the discharge pathways from the reactor building.
{"title":"Research on the refined calculation of released effluent source term from the reactor building in PWR nuclear power plants","authors":"Lu Li, Guoqiang Wei, Junnan Zhang, Jingyun Huang, Qingning Zuo, Shuhan Zhuang, Xiaoping Bai","doi":"10.1016/j.anucene.2025.111270","DOIUrl":"10.1016/j.anucene.2025.111270","url":null,"abstract":"<div><div>The paper takes radionuclide in the reactor building of PWR under normal condition as the research object, a set of nuclide balance equations is established through analyzing the production and disappearance terms of radionuclide in the reactor building. According to the intermittent operation of small sweep gas and introduction of the decay chain, both periodic iteration method and transmutation trajectory analysis method (TTA) are employed to solve for the quantity of radioactive nuclides in the reactor building. Building a virtual nuclear power plant as an example, the radioactive source terms released to the environment from the reactor building is analyzed based on the discharge pathways from the reactor building.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111270"},"PeriodicalIF":1.9,"publicationDate":"2025-02-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143474480","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 : 2025-02-22DOI: 10.1016/j.anucene.2025.111287
Junsen Fu , Yao Xiao , Ziming Wang , Zhengyang Cao , Hanyang Gu
Helical fuel represents an innovative nuclear fuel design characterized by an extended heat transfer surface and reduced fuel temperatures compared to conventional cylindrical rods, significantly enhancing thermal–hydraulic performance. Critical Heat Flux (CHF), a key parameter in light water reactor thermal design, dictates the operational safety margin by marking the threshold of the boiling crisis. However, the flow boiling and CHF characteristics of helical fuel have not been fully elucidated. This study employs high-speed visualization to systematically investigate flow boiling dynamics and CHF characteristics in a single helical fuel rod under atmospheric pressure. During the experiments, visualization measurements are performed to reveal the boiling crisis triggering mechanism. Various flow patterns, including bubbly flow, bubbly-cap flow, slug flow, and annular flow, are identified in steam-water two-phase systems. Our observations indicate that bubble nucleation and aggregation initiate preferentially in the elbow region (the zone of peak heat flux), with subsequent vapor transport along helical blade-induced swirling paths. At elevated steam qualities, localized vapor accumulation and temperature escalation at the elbow heating wall jointly trigger CHF onset. This work examines the effects of thermal–hydraulic and geometric parameters on CHF. The increasing R2/R1 ratio (where R1 is the outer radius of the petal and R2 is the inner radius of the fuel rod) improves CHF due to a more uniform circumferential heat flux. Conversely, a reduction in CHF is observed when the helical pitch decreases from 600 mm to 300 mm, as the enhanced lateral flow promotes vapor phase accumulation. These findings showed that, regardless of the tested geometry, CHF values were about 14.1 % higher than those predicted by the lookup table. The results highlight the importance of optimizing helical fuel geometry to further enhance CHF performance and provide valuable insights for developing advanced safety analysis methods.
{"title":"Visualization experimental investigation on flow boiling and critical heat flux characteristics of helical fuel","authors":"Junsen Fu , Yao Xiao , Ziming Wang , Zhengyang Cao , Hanyang Gu","doi":"10.1016/j.anucene.2025.111287","DOIUrl":"10.1016/j.anucene.2025.111287","url":null,"abstract":"<div><div>Helical fuel represents an innovative nuclear fuel design characterized by an extended heat transfer surface and reduced fuel temperatures compared to conventional cylindrical rods, significantly enhancing thermal–hydraulic performance. Critical Heat Flux (CHF), a key parameter in light water reactor thermal design, dictates the operational safety margin by marking the threshold of the boiling crisis. However, the flow boiling and CHF characteristics of helical fuel have not been fully elucidated. This study employs high-speed visualization to systematically investigate flow boiling dynamics and CHF characteristics in a single helical fuel rod under atmospheric pressure. During the experiments, visualization measurements are performed to reveal the boiling crisis triggering mechanism. Various flow patterns, including bubbly flow, bubbly-cap flow, slug flow, and annular flow, are identified in steam-water two-phase systems. Our observations indicate that bubble nucleation and aggregation initiate preferentially in the elbow region (the zone of peak heat flux), with subsequent vapor transport along helical blade-induced swirling paths. At elevated steam qualities, localized vapor accumulation and temperature escalation at the elbow heating wall jointly trigger CHF onset. This work examines the effects of thermal–hydraulic and geometric parameters on CHF. The increasing <em>R<sub>2</sub>/R<sub>1</sub></em> ratio (where <em>R<sub>1</sub></em> is the outer radius of the petal and <em>R<sub>2</sub></em> is the inner radius of the fuel rod) improves CHF due to a more uniform circumferential heat flux. Conversely, a reduction in CHF is observed when the helical pitch decreases from 600 mm to 300 mm, as the enhanced lateral flow promotes vapor phase accumulation. These findings showed that, regardless of the tested geometry, CHF values were about 14.1 % higher than those predicted by the lookup table. The results highlight the importance of optimizing helical fuel geometry to further enhance CHF performance and provide valuable insights for developing advanced safety analysis methods.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111287"},"PeriodicalIF":1.9,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143464156","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 : 2025-02-22DOI: 10.1016/j.anucene.2025.111260
Inyup Kim, Taesuk Oh, Yonghee Kim
This paper presents the development, optimization, and verification of a depletion module integrated into the iMC Monte Carlo code. Several techniques are implemented to improve the performance and accuracy of the iMC depletion module. In addition, the nuclide control for the depletion of the molten salt reactors is developed. The performance of the depletion module is rigorously assessed through comprehensive code-to-code comparisons with the pre-validated Monte Carlo code Serpent. The evaluation encompasses three distinct depletion scenarios: a single PWR fuel pin, a single SFR fuel pin, and VERA benchmarks. Furthermore, the analysis extends to a simplified molten salt reactor experiment (MSRE) model, incorporating nuclide removal techniques. Comparisons focus on burnup-dependent infinite multiplication factors (kinf) and nuclide densities of actinides and fission products. Results demonstrate both the high accuracy and enhanced efficiency of the iMC Monte Carlo code’s depletion module, marking a significant advancement in advanced reactor analysis capabilities.
{"title":"Development and verification of depletion capabilities in the iMC Monte Carlo code","authors":"Inyup Kim, Taesuk Oh, Yonghee Kim","doi":"10.1016/j.anucene.2025.111260","DOIUrl":"10.1016/j.anucene.2025.111260","url":null,"abstract":"<div><div>This paper presents the development, optimization, and verification of a depletion module integrated into the iMC Monte Carlo code. Several techniques are implemented to improve the performance and accuracy of the iMC depletion module. In addition, the nuclide control for the depletion of the molten salt reactors is developed. The performance of the depletion module is rigorously assessed through comprehensive code-to-code comparisons with the pre-validated Monte Carlo code Serpent. The evaluation encompasses three distinct depletion scenarios: a single PWR fuel pin, a single SFR fuel pin, and VERA benchmarks. Furthermore, the analysis extends to a simplified molten salt reactor experiment (MSRE) model, incorporating nuclide removal techniques. Comparisons focus on burnup-dependent infinite multiplication factors (<em>k<sub>inf</sub></em>) and nuclide densities of actinides and fission products. Results demonstrate both the high accuracy and enhanced efficiency of the iMC Monte Carlo code’s depletion module, marking a significant advancement in advanced reactor analysis capabilities.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111260"},"PeriodicalIF":1.9,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143464155","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111283
Shilei Han , Pengfei Liu , Gang Wang, Bo Kuang, Yanhua Yang
The In-Vessel Retention (IVR) strategy has been widely applied in the existing and newly design Light Water Reactor (LWR). To give a sufficient safety margin for the further design of the large-scale reactors, the enhancement of the Critical Heat Flux (CHF) should be further studied. Porous coating technology is known as an effective CHF enhancement method. In this paper, the REPEC-III facility is modified to adapt the CHF enhancement conditions with higher heating power and wall temperature. The REPEC-III facility has 1:1 height ratio with the prototypic External Reactor Vessel Cooling (ERVC) environment. The flow channel is designed as a curved rectangular channel and the area ratio is 1:100 to the prototypic ERVC flow channel. The applicability of the porous coatings under the IVR-ERVC conditions are evaluated in this study. The cold spray technology is applied to prepare the porous coatings and prevent surface damage. The porous layer is composed of the dense basal layer and porous layer. The comparisons of the boiling phenomena are analyzed. On the porous surface, the larger amplitude and lower frequency of the pressure difference oscillation are observed and mean the more vapor slugs and the longer vapor period. During the experiments, the temperatures of the heating block with the porous surface are overall higher than the temperatures with the fresh surface. The heat transfer capability is worsened by the thermal resistance and improved liquid replenishment, which leads to a higher wall superheat under the same heat flux. The CHF is enhanced by the porous coatings and the enhancement effect is related to the orientations. The maximum percentage of the CHF enhancement is 43% at 87°. The intense turbulent induced by the more vapor slugs and the capillary wicking in the porous layer are beneficial to the liquid replenishment and enhance the CHF. Through the full-height experimental analysis, the porous coating technology is an effective method to improve the safety margin of the IVR strategy. The application of the porous coatings on the Reactor Pressure Vessel (RPV) outer wall are still needed to be further studied.
{"title":"Nucleate boiling heat transfer and CHF enhancement with porous surface coatings on the RPV outer wall","authors":"Shilei Han , Pengfei Liu , Gang Wang, Bo Kuang, Yanhua Yang","doi":"10.1016/j.anucene.2025.111283","DOIUrl":"10.1016/j.anucene.2025.111283","url":null,"abstract":"<div><div>The In-Vessel Retention (IVR) strategy has been widely applied in the existing and newly design Light Water Reactor (LWR). To give a sufficient safety margin for the further design of the large-scale reactors, the enhancement of the Critical Heat Flux (CHF) should be further studied. Porous coating technology is known as an effective CHF enhancement method. In this paper, the REPEC-III facility is modified to adapt the CHF enhancement conditions with higher heating power and wall temperature. The REPEC-III facility has 1:1 height ratio with the prototypic External Reactor Vessel Cooling (ERVC) environment. The flow channel is designed as a curved rectangular channel and the area ratio is 1:100 to the prototypic ERVC flow channel. The applicability of the porous coatings under the IVR-ERVC conditions are evaluated in this study. The cold spray technology is applied to prepare the porous coatings and prevent surface damage. The porous layer is composed of the dense basal layer and porous layer. The comparisons of the boiling phenomena are analyzed. On the porous surface, the larger amplitude and lower frequency of the pressure difference oscillation are observed and mean the more vapor slugs and the longer vapor period. During the experiments, the temperatures of the heating block with the porous surface are overall higher than the temperatures with the fresh surface. The heat transfer capability is worsened by the thermal resistance and improved liquid replenishment, which leads to a higher wall superheat under the same heat flux. The CHF is enhanced by the porous coatings and the enhancement effect is related to the orientations. The maximum percentage of the CHF enhancement is 43% at 87°. The intense turbulent induced by the more vapor slugs and the capillary wicking in the porous layer are beneficial to the liquid replenishment and enhance the CHF. Through the full-height experimental analysis, the porous coating technology is an effective method to improve the safety margin of the IVR strategy. The application of the porous coatings on the Reactor Pressure Vessel (RPV) outer wall are still needed to be further studied.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111283"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454360","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111249
Laurent Laborde, Benoît Migot
Loss of cooling in a Spent Fuel Pool (SFP) of a nuclear power plant can lead to the melting of fuel assemblies and to strong radiological consequences to the environment. In order to study the first phases of such accidents, up to the fuel assemblies uncovery, the DENOPI project was launched by the French Institute for Radiation Protection and Nuclear Safety (IRSN1) supported and funded by the French Government and partners. Among the different facilities developed in the project, the MIDI facility aims at studying the complex thermal-hydraulics phenomena occurring in a large water pool heated from the bottom by electrical rods arranged in dedicated racks. MIDI is scaled by homothety to a typical French SFP. Different assembly arrangements (loading patterns) have been tested at different power levels, with either uniform power repartition, or hot and cold cells. In each test, the water level and temperatures at different elevations are followed, as well as mass flow rate entering each fuel rack. These experimental results also provide relevant data for the analysis and understanding of large natural convection loops that are expected in immersed passive heat removal systems of Small Modular Reactors. The forthcoming OECD/NEA POLCA project aims to extend such results database, in particular to assess the capability of thermal-hydraulics codes to reproduce the main tendencies of these experimental results.
The ASTEC code developed by IRSN is a system code dedicated to the simulation of major accidents in nuclear facilities that may lead to the release of radiological material. Recent works within the MUSA European project have shown the importance of reducing models uncertainties in the first phases of the accident, during the pool dewatering. In this paper, first simulations of MIDI tests are performed with ASTEC in order to assess and improve the capability of ASTEC to simulate the dewatering of a large water pool such as a SFP during a loss-of-cooling accident. Simulations are performed for a selection of MIDI tests with different heating patterns and power levels. Different models of subcooled boiling models from the literature are tested in ASTEC, stressing the key role of these models for an accurate prediction of the experimental flow.
{"title":"ASTEC validation of SFP dewatering using results from the DENOPI project","authors":"Laurent Laborde, Benoît Migot","doi":"10.1016/j.anucene.2025.111249","DOIUrl":"10.1016/j.anucene.2025.111249","url":null,"abstract":"<div><div>Loss of cooling in a Spent Fuel Pool (SFP) of a nuclear power plant can lead to the melting of fuel assemblies and to strong radiological consequences to the environment. In order to study the first phases of such accidents, up to the fuel assemblies uncovery, the DENOPI project was launched by the French Institute for Radiation Protection and Nuclear Safety (IRSN<span><span><sup>1</sup></span></span>) supported and funded by the French Government and partners. Among the different facilities developed in the project, the MIDI facility aims at studying the complex thermal-hydraulics phenomena occurring in a large water pool heated from the bottom by electrical rods arranged in dedicated racks. MIDI is scaled by homothety to a typical French SFP. Different assembly arrangements (loading patterns) have been tested at different power levels, with either uniform power repartition, or hot and cold cells. In each test, the water level and temperatures at different elevations are followed, as well as mass flow rate entering each fuel rack. These experimental results also provide relevant data for the analysis and understanding of large natural convection loops that are expected in immersed passive heat removal systems of Small Modular Reactors. The forthcoming OECD/NEA POLCA project aims to extend such results database, in particular to assess the capability of thermal-hydraulics codes to reproduce the main tendencies of these experimental results.</div><div>The ASTEC code developed by IRSN is a system code dedicated to the simulation of major accidents in nuclear facilities that may lead to the release of radiological material. Recent works within the MUSA European project have shown the importance of reducing models uncertainties in the first phases of the accident, during the pool dewatering. In this paper, first simulations of MIDI tests are performed with ASTEC in order to assess and improve the capability of ASTEC to simulate the dewatering of a large water pool such as a SFP during a loss-of-cooling accident. Simulations are performed for a selection of MIDI tests with different heating patterns and power levels. Different models of subcooled boiling models from the literature are tested in ASTEC, stressing the key role of these models for an accurate prediction of the experimental flow.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111249"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143464154","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111286
Long Yun, Xu Yuan, Guo Xi’an, Zhang Mingyu
In this paper, proper orthogonal decomposition (POD) technique is used to analyze the internal flow excitation characteristics of the reactor coolant pump. The stable operation of the reactor coolant pump (RCP), a critical component of nuclear power plants, is essential for maintaining reactor core cooling. The influence of the lower chamber of the steam generator on the pump inlet conditions is considered. Through numerical simulation and feature extraction techniques, the flow patterns and dynamic behaviors of key components such as impeller, diffuser and casing are analyzed in depth, and the multi-transient data of RCP are successfully processed. The POD analysis identifies the dominant energy structures within the flow field, offering insights into the primary flow characteristics. Studies have shown that POD technology can not only identify and explain complex flow phenomena under non-uniform inflow conditions, but also significantly improve the performance improvement and fault prevention capabilities of reactor coolant pumps.
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