Pub Date : 2025-11-25DOI: 10.1016/j.fusengdes.2025.115540
Bin Yu , Yongtao An , Dong Wang , Guangcheng Yang , Fei Gao , Xueling Zhang , Fei Jiang , Min Chen , Xin Zhang , Zhanghong Shi , Can Tang , Ang Song , Dongwen Wang , Jiangfeng Song , Yan Shi , Changan Chen , Wei Shi , Peilong Li , Wenhua Luo
As nuclear and fusion energy technologies advance, the tritiated water management poses a critical challenge. Herein, we developed and constructed a heat-pump-integrated water distillation facility, which enables energy-efficient and high-performance treatment of tritiated water. This facility achieved a peak tritium Detritiation Factor (DF) of (2.72 ± 0.16) × 104 during continuous processing at feedwater throughputs of 10.6 L/h, reducing the tritium activity from 1.82 × 105 Bq/L to 6.7 Bq/L. An operational assessment with a maxim um processing capacity of 20.6 L/h has been carried out and DF of 315 was achieved. Furthermore, our developed structured packing DTC-APD demonstrated a height equivalent of a theoretical plate (HETP) within the range of 9.9–10.6 cm, surpassing performance levels typically reported in the literature for similar structured packings in tritiated water treatment facility. Notably, the energy consumption for processing tritiated wastewater decreased by 77.7% compared to conventional distillation techniques. Heat pump distillation was successfully applied to tritiated water treatment, maintaining stable operation for 1632 h. Therefore, this method enables economical and efficient processing of large-volume, low-concentration tritiated wastewater, which represents a significant advancement for tritiated water engineering in nuclear and fusion energy applications.
{"title":"An heat-pump-integrated water distillation facility for high-efficiency, ultralow energy consumption tritiated water treatment","authors":"Bin Yu , Yongtao An , Dong Wang , Guangcheng Yang , Fei Gao , Xueling Zhang , Fei Jiang , Min Chen , Xin Zhang , Zhanghong Shi , Can Tang , Ang Song , Dongwen Wang , Jiangfeng Song , Yan Shi , Changan Chen , Wei Shi , Peilong Li , Wenhua Luo","doi":"10.1016/j.fusengdes.2025.115540","DOIUrl":"10.1016/j.fusengdes.2025.115540","url":null,"abstract":"<div><div>As nuclear and fusion energy technologies advance, the tritiated water management poses a critical challenge. Herein, we developed and constructed a heat-pump-integrated water distillation facility, which enables energy-efficient and high-performance treatment of tritiated water. This facility achieved a peak tritium Detritiation Factor (<em>DF</em>) of (2.72 ± 0.16) × 10<sup>4</sup> during continuous processing at feedwater throughputs of 10.6 L/h, reducing the tritium activity from 1.82 × 10<sup>5</sup> Bq/L to 6.7 Bq/L. An operational assessment with a maxim um processing capacity of 20.6 L/h has been carried out and <em>DF</em> of 315 was achieved. Furthermore, our developed structured packing DTC-APD demonstrated a height equivalent of a theoretical plate (<em>HETP</em>) within the range of 9.9–10.6 cm, surpassing performance levels typically reported in the literature for similar structured packings in tritiated water treatment facility. Notably, the energy consumption for processing tritiated wastewater decreased by 77.7% compared to conventional distillation techniques. Heat pump distillation was successfully applied to tritiated water treatment, maintaining stable operation for 1632 h. Therefore, this method enables economical and efficient processing of large-volume, low-concentration tritiated wastewater, which represents a significant advancement for tritiated water engineering in nuclear and fusion energy applications.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115540"},"PeriodicalIF":2.0,"publicationDate":"2025-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684654","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}
Low-energy and high-flux helium plasma irradiation can cause serious erosion and damage to the surface of tungsten (W), resulting in the formation of defects, holes and fuzz structures. The formation of these surface structures will significantly degrade the material's thermal conductivity, disrupt surface temperature distribution, and lead to surface degradation, thereby reducing the irradiation resistance of the material.To address this issue, this study introduces lanthanum oxide (La2O3) as a second phase to effectively reduce large-angle grain boundaries on the W surface and inhibit helium bubble formation by preventing helium plasma accumulation. The result shows that compared with pure tungsten (PW), W-La2O3 composite (the mass fraction of La2O3 second phase is 1 %) exhibits stronger resistance to helium plasma irradiation. Under the irradiation condition of helium plasma beam current of 2.91×1021 ions/m2·s and the dose of 3.492×1024 ions/m2, the W-La2O3 composite surface presents a wavy surface structure, which is different from the typical fuzzy surface structure formed on the pure W, indicating a significant alteration in defect evolution. With the irradiation dose increased to 13.020×1024 ions/m2, the wavy surface structure disappears, and the surface structure exhibits a classical pyramidal surface structure. The simulation results of SRIM software further reveal that the La2O3 second phase helps to transfer the helium plasma gathered on the W surface to a deeper place inside the material. In this way, the aggregation of helium ions on the surface is effectively alleviated, the formation process of the fuzz structure is delayed, and the stability of the irradiated surface structure is significantly improved. This study provides a novel design strategy for improving the anti-radiation performance of W-based plasma-facing materials in extreme fusion environments.
{"title":"The effect of La2O3 on the irradiation resistance of tungsten under low energy and high flux helium plasma irradiation","authors":"Kaichao Fu , Dang Xu , Changcheng Sang , Ruizhi Chen , Pengqi Chen , Dahuan Zhu , Qiu Xu , Jigui Cheng","doi":"10.1016/j.fusengdes.2025.115531","DOIUrl":"10.1016/j.fusengdes.2025.115531","url":null,"abstract":"<div><div>Low-energy and high-flux helium plasma irradiation can cause serious erosion and damage to the surface of tungsten (W), resulting in the formation of defects, holes and fuzz structures. The formation of these surface structures will significantly degrade the material's thermal conductivity, disrupt surface temperature distribution, and lead to surface degradation, thereby reducing the irradiation resistance of the material.To address this issue, this study introduces lanthanum oxide (La<sub>2</sub>O<sub>3</sub>) as a second phase to effectively reduce large-angle grain boundaries on the W surface and inhibit helium bubble formation by preventing helium plasma accumulation. The result shows that compared with pure tungsten (PW), W-La<sub>2</sub>O<sub>3</sub> composite (the mass fraction of La<sub>2</sub>O<sub>3</sub> second phase is 1 %) exhibits stronger resistance to helium plasma irradiation. Under the irradiation condition of helium plasma beam current of 2.91×10<sup>21</sup> ions/m<sup>2</sup>·s and the dose of 3.492×10<sup>24</sup> ions/m<sup>2</sup>, the W-La<sub>2</sub>O<sub>3</sub> composite surface presents a wavy surface structure, which is different from the typical fuzzy surface structure formed on the pure W, indicating a significant alteration in defect evolution. With the irradiation dose increased to 13.020×10<sup>24</sup> ions/m<sup>2</sup>, the wavy surface structure disappears, and the surface structure exhibits a classical pyramidal surface structure. The simulation results of SRIM software further reveal that the La<sub>2</sub>O<sub>3</sub> second phase helps to transfer the helium plasma gathered on the W surface to a deeper place inside the material. In this way, the aggregation of helium ions on the surface is effectively alleviated, the formation process of the fuzz structure is delayed, and the stability of the irradiated surface structure is significantly improved. This study provides a novel design strategy for improving the anti-radiation performance of W-based plasma-facing materials in extreme fusion environments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115531"},"PeriodicalIF":2.0,"publicationDate":"2025-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684651","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}
Reduced-activation ferritic/martensitic steel (F82H) is considered as a promising structural material for the fusion blanket structural system following the DEMO reactor. For blanket fabrication, hot isostatic pressing (HIP) is employed for the production of components with complex geometries. A key challenge associated with the HIP processing of F82H steels is the formation of silicon oxide at the HIPed interface, which results in reduced joint toughness. This study investigates the correlation between the thermodynamic stability of oxides and researches whether the silicon precipitation is the special property of silicon or not. An FeCrAl alloy containing Si was selected for comparison with F82H steel, and both materials were subjected to identical processing conditions. Silicon oxides were observed at the F82H steel-HIPed interface, whereas aluminum oxides were observed in the FeCrAl alloy. No Si precipitation or accumulation was detected at the HIP-treated interface of the FeCrAl alloy. This research showed that the silicon precipitation on F82H HIPed interface is not a special property of silicon but a general behavior of the elements along the thermodynamics and formation energy of oxides.
{"title":"Study of chemical content effects for oxide formation on HIPed interface during the fabrication process by comparison of F82H steel with FeCrAl alloy","authors":"Hirotatsu Kishimoto , Tamaki Shibayama , Takashi Nozawa , Hiroyasu Tanigawa","doi":"10.1016/j.fusengdes.2025.115541","DOIUrl":"10.1016/j.fusengdes.2025.115541","url":null,"abstract":"<div><div>Reduced-activation ferritic/martensitic steel (F82H) is considered as a promising structural material for the fusion blanket structural system following the DEMO reactor. For blanket fabrication, hot isostatic pressing (HIP) is employed for the production of components with complex geometries. A key challenge associated with the HIP processing of F82H steels is the formation of silicon oxide at the HIPed interface, which results in reduced joint toughness. This study investigates the correlation between the thermodynamic stability of oxides and researches whether the silicon precipitation is the special property of silicon or not. An FeCrAl alloy containing Si was selected for comparison with F82H steel, and both materials were subjected to identical processing conditions. Silicon oxides were observed at the F82H steel-HIPed interface, whereas aluminum oxides were observed in the FeCrAl alloy. No Si precipitation or accumulation was detected at the HIP-treated interface of the FeCrAl alloy. This research showed that the silicon precipitation on F82H HIPed interface is not a special property of silicon but a general behavior of the elements along the thermodynamics and formation energy of oxides.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115541"},"PeriodicalIF":2.0,"publicationDate":"2025-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684653","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 : 2025-11-24DOI: 10.1016/j.fusengdes.2025.115533
Wei Zheng , Qian Xu , Jichan Xu , Xin Yang , Haishan Zhou , Guangnan Luo
To meet the demanding requirements for high-precision, spatially-resolved diagnostics in plasma-material interaction (PMI) studies under the high-flux, high-magnetic-field environment of the SPARROW linear plasma device, an actively water-cooled Langmuir probe array system has been designed and developed. This design synergizes actively water-cooling with array layout requirements, with a focus on optimizing the cooling channel structure. Through systematic computational fluid dynamics (CFD) simulations, the thermal performance of the probe was quantitatively evaluated under Gaussian-distributed heat fluxes of 10 MW/m², 15 MW/m², and 20 MW/m², along with the impact on the probe body and the key insulating material (alumina ceramic). Under the 15 MW/m² heat flux, the maximum temperatures of the tungsten tip and alumina sleeve are maintained at approximately 62% and 61% of their respective safety limits. Even under the extreme 20 MW/m² condition, these key diagnostic components remain below 75% of their limits, demonstrating a substantial safety buffer that accommodates potential CFD uncertainties. By integrating innovative design with comprehensive thermal analysis, this research establishes key technical foundations for achieving efficient and reliable arrayed active diagnostics in extreme fusion-relevant plasma environments. It provides vital support for future high-parameter plasma physics experiments.
{"title":"Design and thermal analysis of an actively water-cooled array probe for the SPARROW device","authors":"Wei Zheng , Qian Xu , Jichan Xu , Xin Yang , Haishan Zhou , Guangnan Luo","doi":"10.1016/j.fusengdes.2025.115533","DOIUrl":"10.1016/j.fusengdes.2025.115533","url":null,"abstract":"<div><div>To meet the demanding requirements for high-precision, spatially-resolved diagnostics in plasma-material interaction (PMI) studies under the high-flux, high-magnetic-field environment of the SPARROW linear plasma device, an actively water-cooled Langmuir probe array system has been designed and developed. This design synergizes actively water-cooling with array layout requirements, with a focus on optimizing the cooling channel structure. Through systematic computational fluid dynamics (CFD) simulations, the thermal performance of the probe was quantitatively evaluated under Gaussian-distributed heat fluxes of 10 MW/m², 15 MW/m², and 20 MW/m², along with the impact on the probe body and the key insulating material (alumina ceramic). Under the 15 MW/m² heat flux, the maximum temperatures of the tungsten tip and alumina sleeve are maintained at approximately 62% and 61% of their respective safety limits. Even under the extreme 20 MW/m² condition, these key diagnostic components remain below 75% of their limits, demonstrating a substantial safety buffer that accommodates potential CFD uncertainties. By integrating innovative design with comprehensive thermal analysis, this research establishes key technical foundations for achieving efficient and reliable arrayed active diagnostics in extreme fusion-relevant plasma environments. It provides vital support for future high-parameter plasma physics experiments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115533"},"PeriodicalIF":2.0,"publicationDate":"2025-11-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684658","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 : 2025-11-22DOI: 10.1016/j.fusengdes.2025.115530
Tamás Szepesi , A. Buzás , G. Cseh , G. Kocsis , Á. Kovácsik , D.I. Réfy , T. Szabolics , H. Homma , T. Nakano , M. Yoshida , W. Bin , G. De Tommasi , F. Fiorenza , D. Frattolillo , M. Iafrati , M. Mattei , A. Pironti , D. Ricci , C. Sozzi , J. Svoboda , J. Cavalier
EDICAM (Event Detection Intelligent Camera), a wide-angle visible overview camera diagnostic, was operated throughout the successful Integrated Commissioning and first operation campaign (IC&OP1) of JT-60SA, the world’s largest superconducting tokamak. It provided immediate and essential visual feedback after each discharge, supporting plasma operation and scenario development. In early attempts where burn-through failed and plasma current remained low, EDICAM captured plasmas appearing as two thin, radiative cylindrical surfaces—matching the resonance layers of the applied electron cyclotron heating waves. During successful discharges, the formation of magnetic surfaces was confirmed. Smaller plasmas, partially filling the torus, showed a distinct radiation belt at the low-temperature plasma edge. This belt served as a visual proxy for tracking plasma size evolution. For larger plasmas, EDICAM images were dominated by bright regions of plasma-wall interaction (PWI), especially at the limiters and the central column’s heat shield. The transition from a visible radiation belt to pronounced PWI regions indicated that the plasma had reached its maximum size. Diverted plasma formation was also identified by the presence of divertor strike-lines, which brightened as the plasma current increased. These strike-lines gradually shifted inward, consistent with the decreasing current in the central solenoid due to flux consumption. In addition, EDICAM detected unexpected PWI events, such as hot-spots, which resulted in sprays of particles entering the plasma. Altogether, EDICAM proved to be an essential diagnostic tool for the interpretation of plasma behavior during JT-60SA’s initial operational phase.
{"title":"Analysis of the first plasmas of JT-60SA using the EDICAM visible video diagnostic","authors":"Tamás Szepesi , A. Buzás , G. Cseh , G. Kocsis , Á. Kovácsik , D.I. Réfy , T. Szabolics , H. Homma , T. Nakano , M. Yoshida , W. Bin , G. De Tommasi , F. Fiorenza , D. Frattolillo , M. Iafrati , M. Mattei , A. Pironti , D. Ricci , C. Sozzi , J. Svoboda , J. Cavalier","doi":"10.1016/j.fusengdes.2025.115530","DOIUrl":"10.1016/j.fusengdes.2025.115530","url":null,"abstract":"<div><div>EDICAM (Event Detection Intelligent Camera), a wide-angle visible overview camera diagnostic, was operated throughout the successful Integrated Commissioning and first operation campaign (IC&OP1) of JT-60SA, the world’s largest superconducting tokamak. It provided immediate and essential visual feedback after each discharge, supporting plasma operation and scenario development. In early attempts where burn-through failed and plasma current remained low, EDICAM captured plasmas appearing as two thin, radiative cylindrical surfaces—matching the resonance layers of the applied electron cyclotron heating waves. During successful discharges, the formation of magnetic surfaces was confirmed. Smaller plasmas, partially filling the torus, showed a distinct radiation belt at the low-temperature plasma edge. This belt served as a visual proxy for tracking plasma size evolution. For larger plasmas, EDICAM images were dominated by bright regions of plasma-wall interaction (PWI), especially at the limiters and the central column’s heat shield. The transition from a visible radiation belt to pronounced PWI regions indicated that the plasma had reached its maximum size. Diverted plasma formation was also identified by the presence of divertor strike-lines, which brightened as the plasma current increased. These strike-lines gradually shifted inward, consistent with the decreasing current in the central solenoid due to flux consumption. In addition, EDICAM detected unexpected PWI events, such as hot-spots, which resulted in sprays of particles entering the plasma. Altogether, EDICAM proved to be an essential diagnostic tool for the interpretation of plasma behavior during JT-60SA’s initial operational phase.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115530"},"PeriodicalIF":2.0,"publicationDate":"2025-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684566","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 : 2025-11-22DOI: 10.1016/j.fusengdes.2025.115532
Kok Yue Chan , Pengfei Shen , Zhaoyuan Liu , Songlin Liu , Zhe Chuan Tan , Yuanhao Gou , Shihang Jiang , Kan Wang
This study explores the application of the Reactor Monte Carlo (RMC) code in fusion neutronics, demonstrating its capabilities in various areas of fusion reactor analysis. We developed the MCNP-to-RMC conversion tool (M2R) to facilitate the transition from MCNP to RMC format, validated by converting the ITER C-model and CFETR model. The study further investigates the use of RMC for global flux distribution calculations in the CFETR model, employing a density reduction method for weight window generation, showcasing RMC’s variance reduction and neutron transport capabilities in complex fusion reactor geometries. Additionally, the Tritium Breeding Ratio (TBR) for the CFETR blanket was calculated and compared with OpenMC, showing excellent agreement. The study also assessed RMC’s CAD-based transport capabilities using the Paramak to model a half-tokamak design, with simulations performed in both RMC and OpenMC, showing consistent results. Overall, the findings validate RMC as a reliable tool for fusion neutronics simulations, demonstrating its capability to handle complex reactor models, neutron transport, and key parameter calculations.
{"title":"Fusion neutronics simulations with RMC: Key capabilities and techniques","authors":"Kok Yue Chan , Pengfei Shen , Zhaoyuan Liu , Songlin Liu , Zhe Chuan Tan , Yuanhao Gou , Shihang Jiang , Kan Wang","doi":"10.1016/j.fusengdes.2025.115532","DOIUrl":"10.1016/j.fusengdes.2025.115532","url":null,"abstract":"<div><div>This study explores the application of the Reactor Monte Carlo (RMC) code in fusion neutronics, demonstrating its capabilities in various areas of fusion reactor analysis. We developed the MCNP-to-RMC conversion tool (M2R) to facilitate the transition from MCNP to RMC format, validated by converting the ITER C-model and CFETR model. The study further investigates the use of RMC for global flux distribution calculations in the CFETR model, employing a density reduction method for weight window generation, showcasing RMC’s variance reduction and neutron transport capabilities in complex fusion reactor geometries. Additionally, the Tritium Breeding Ratio (TBR) for the CFETR blanket was calculated and compared with OpenMC, showing excellent agreement. The study also assessed RMC’s CAD-based transport capabilities using the Paramak to model a half-tokamak design, with simulations performed in both RMC and OpenMC, showing consistent results. Overall, the findings validate RMC as a reliable tool for fusion neutronics simulations, demonstrating its capability to handle complex reactor models, neutron transport, and key parameter calculations.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115532"},"PeriodicalIF":2.0,"publicationDate":"2025-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684659","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 : 2025-11-19DOI: 10.1016/j.fusengdes.2025.115517
Andrea Burlando , Luciana Barucca , Fabio Giannetti , Cristiano Ciurluini , Francesco Colliva , Giorgio Mongiardini , Carlo Bima , Alberto Traverso , Amelia Tincani , Marco Mantero
European DEMOnstration fusion power plant (EU-DEMO) has the main goal of demonstrating the possibility of producing hundreds of MWe of electrical power from fusion by the end of the century. In this sense an important role is played by the Balance of Plant (BoP). BoP includes, at least, the Power Conversion System (PCS) that converts thermal power into electricity. PCS is thermally connected to the Primary Heat Transfer System (PHTS) which collects the thermal power generated in the reactor. In case of DEMO equipped by a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), three different architectures of BoP are under development. In this work will be analyzed the Indirect BoP architecture, so called ICD (Indirect Concept Design) PCS, characterized by an Intermediate Heat Transport System (IHTS) installed between PHTS and PCS and equipped with a Small Energy Storage (molten salt) to operate, during Dwell, the Steam Turbine at around 10 % nominal steam load, as in the reference cycle. In this work it will be highlighted: i) the PCS layout and the relative Heat&Mass balance, ii) the cycle performance analysis and iii) the performance comparison with the other “pulsed” PCS architectures above mentioned. Some hints on the concept design of the interfacing Intermediate Heat Transfer Circuit with Small Energy Storage will be also provided.
{"title":"Water cooled lithium lead balance of plant indirectly coupled with power conversion system operated with small energy storage in EU DEMO","authors":"Andrea Burlando , Luciana Barucca , Fabio Giannetti , Cristiano Ciurluini , Francesco Colliva , Giorgio Mongiardini , Carlo Bima , Alberto Traverso , Amelia Tincani , Marco Mantero","doi":"10.1016/j.fusengdes.2025.115517","DOIUrl":"10.1016/j.fusengdes.2025.115517","url":null,"abstract":"<div><div>European DEMOnstration fusion power plant (EU-DEMO) has the main goal of demonstrating the possibility of producing hundreds of MWe of electrical power from fusion by the end of the century. In this sense an important role is played by the Balance of Plant (BoP). BoP includes, at least, the Power Conversion System (PCS) that converts thermal power into electricity. PCS is thermally connected to the Primary Heat Transfer System (PHTS) which collects the thermal power generated in the reactor. In case of DEMO equipped by a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), three different architectures of BoP are under development. In this work will be analyzed the Indirect BoP architecture, so called ICD (Indirect Concept Design) PCS, characterized by an Intermediate Heat Transport System (IHTS) installed between PHTS and PCS and equipped with a Small Energy Storage (molten salt) to operate, during Dwell, the Steam Turbine at around 10 % nominal steam load, as in the reference cycle. In this work it will be highlighted: i) the PCS layout and the relative Heat&Mass balance, ii) the cycle performance analysis and iii) the performance comparison with the other “pulsed” PCS architectures above mentioned. Some hints on the concept design of the interfacing Intermediate Heat Transfer Circuit with Small Energy Storage will be also provided.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115517"},"PeriodicalIF":2.0,"publicationDate":"2025-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684660","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The ADITYA tokamak (R₀ = 0.75 m, a = 0.25 m), originally operated with a limiter configuration, has been successfully upgraded to ADITYA-U with an open divertor configuration to enhance plasma confinement and operational flexibility. Limiter and divertor systems define the plasma boundary within the vacuum vessel, protecting in-vessel components by minimizing direct plasma-wall interactions. In both single-null and double-null divertor configurations, ADITYA-U is designed to produce circular and shaped plasmas with a triangularity (δ) of ∼0.45, elongation (κ) of ∼1.1–1.2, and plasma currents (Ip) in the range of 100–150 kA.
Using the plasma equilibrium simulation code IPREQ, the optimal locations for limiter and divertor plates were determined and validated for the new vacuum vessel geometry. Plasma-facing components (PFCs) based on graphite—including toroidal, poloidal, and safety limiters, as well as divertor tiles—were installed in a staged manner to facilitate a progressive operational strategy. Limiters were installed prior to the initial operation phase to manage plasma-wall interactions during early campaigns, while divertor plates were added subsequently, following operational experience with impurity behavior during the burn-through phase.
This paper details the preparatory studies, simulation-guided design process, and the in-situ installation challenges associated with retrofitting graphite limiter and divertor assemblies in the ADITYA-U tokamak. The work provides insights into the phased upgrade process of a medium-sized tokamak and highlights practical strategies for integrating advanced plasma boundary configurations in existing devices.
ADITYA托卡马克（R 0 = 0.75 m, a = 0.25 m）最初采用限流器配置，现已成功升级为ADITYA- u，采用开放式导流器配置，以增强等离子体约束和操作灵活性。限制器和分流器系统定义了真空容器内的等离子体边界，通过最大限度地减少等离子体与容器壁的直接相互作用来保护容器内的组件。在单零和双零分流器配置中，ADITYA-U设计用于产生三角形（δ）为~ 0.45，伸长率（κ）为~ 1.1-1.2的圆形和形状等离子体，等离子体电流（Ip）在100-150 kA范围内。利用等离子体平衡模拟代码IPREQ，确定了限制板和导流板的最佳位置，并对新真空容器的几何形状进行了验证。基于石墨的等离子体组件（pfc），包括环向、极向、安全限制器以及导流砖，都是分阶段安装的，以促进渐进式作业策略。在初始操作阶段之前安装了限位器，以管理早期活动期间等离子体壁的相互作用，随后根据在烧透阶段的杂质行为的操作经验添加了导流板。本文详细介绍了ADITYA-U托卡马克中石墨限位器和导流器组件改造的前期研究、仿真指导设计过程以及现场安装挑战。这项工作为中型托卡马克的分阶段升级过程提供了见解，并强调了在现有设备中集成先进等离子体边界配置的实用策略。
{"title":"Design, construction, integration and installation of plasma facing components of limiter & divertor of ADITYA-U tokamak","authors":"K.M. Patel , K.A. Jadeja , Harshita Raj , J. Ghosh , S.B. Bhatt , R.L. Tanna , Deepti Sharma , Arun Prakash , Rupesh G , Suman Aich , Komal Yadav , SK Injamul Hoque , Rohit Kumar , Aditya-U Team","doi":"10.1016/j.fusengdes.2025.115520","DOIUrl":"10.1016/j.fusengdes.2025.115520","url":null,"abstract":"<div><div>The ADITYA tokamak (R₀ = 0.75 m, <em>a</em> = 0.25 m), originally operated with a limiter configuration, has been successfully upgraded to ADITYA-U with an open divertor configuration to enhance plasma confinement and operational flexibility. Limiter and divertor systems define the plasma boundary within the vacuum vessel, protecting in-vessel components by minimizing direct plasma-wall interactions. In both single-null and double-null divertor configurations, ADITYA-U is designed to produce circular and shaped plasmas with a triangularity (δ) of ∼0.45, elongation (κ) of ∼1.1–1.2, and plasma currents (Ip) in the range of 100–150 kA.</div><div>Using the plasma equilibrium simulation code IPREQ, the optimal locations for limiter and divertor plates were determined and validated for the new vacuum vessel geometry. Plasma-facing components (PFCs) based on graphite—including toroidal, poloidal, and safety limiters, as well as divertor tiles—were installed in a staged manner to facilitate a progressive operational strategy. Limiters were installed prior to the initial operation phase to manage plasma-wall interactions during early campaigns, while divertor plates were added subsequently, following operational experience with impurity behavior during the burn-through phase.</div><div>This paper details the preparatory studies, simulation-guided design process, and the in-situ installation challenges associated with retrofitting graphite limiter and divertor assemblies in the ADITYA-U tokamak. The work provides insights into the phased upgrade process of a medium-sized tokamak and highlights practical strategies for integrating advanced plasma boundary configurations in existing devices.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115520"},"PeriodicalIF":2.0,"publicationDate":"2025-11-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145519558","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 : 2025-11-14DOI: 10.1016/j.fusengdes.2025.115522
Yue Yu , Wenjia Wang , Kecheng Jiang , Lei Chen , Songlin Liu
Within the framework of the Comprehensive Research Facility for Fusion Technology (CRAFT) Program of China, the China LIthium-lead Magnetohydrodynamics Blanket expERiment (CLIMBER) facility is currently being constructed. The facility is designed to systematically investigate the intrinsic mechanisms of the MagnetoHydroDynamics (MHD) effect through a series of experiments, thereby providing experimental evidence and theoretical support for optimizing the design of the supercritical carbon dioxide (s-CO2) cOoled Lithium-Lead (COOL) blanket. In this paper, the thermal hydraulic system analysis is performed for CLIMBER facility using the modified RELAP5/MOD3.3. Firstly, the component-level simulations are performed using both ANSYS‑CFX and RELAP5. The accuracy of the modified RELAP5 code in heat transfer calculations is evaluated, and necessary corrections are applied to the pressure loss based on the CFX results. Subsequently, the verified RELAP5 component models are integrated into the system model, which comprises of the PbLi, thermal oil, and water systems, to conduct system‐level analysis. Various operating conditions required for the experiments are simulated by adjusting critical parameters such as valve openings, pump flow rates, and electric heating power. Temperature, pressure, and flow rate at different nodes are obtained, and the maximum pressure drop of each system is calculated. The results not only confirm the feasibility of the loop design, but also provide essential guidance and data support for future experimental implementation and the formulation of operational control strategies.
{"title":"Thermal–hydraulic system analysis of the CLIMBER facility using RELAP5/MOD3.3","authors":"Yue Yu , Wenjia Wang , Kecheng Jiang , Lei Chen , Songlin Liu","doi":"10.1016/j.fusengdes.2025.115522","DOIUrl":"10.1016/j.fusengdes.2025.115522","url":null,"abstract":"<div><div>Within the framework of the Comprehensive Research Facility for Fusion Technology (CRAFT) Program of China, the China LIthium-lead Magnetohydrodynamics Blanket expERiment (CLIMBER) facility is currently being constructed. The facility is designed to systematically investigate the intrinsic mechanisms of the MagnetoHydroDynamics (MHD) effect through a series of experiments, thereby providing experimental evidence and theoretical support for optimizing the design of the supercritical carbon dioxide (s-CO<sub>2</sub>) cOoled Lithium-Lead (COOL) blanket. In this paper, the thermal hydraulic system analysis is performed for CLIMBER facility using the modified RELAP5/MOD3.3. Firstly, the component-level simulations are performed using both ANSYS‑CFX and RELAP5. The accuracy of the modified RELAP5 code in heat transfer calculations is evaluated, and necessary corrections are applied to the pressure loss based on the CFX results. Subsequently, the verified RELAP5 component models are integrated into the system model, which comprises of the PbLi, thermal oil, and water systems, to conduct system‐level analysis. Various operating conditions required for the experiments are simulated by adjusting critical parameters such as valve openings, pump flow rates, and electric heating power. Temperature, pressure, and flow rate at different nodes are obtained, and the maximum pressure drop of each system is calculated. The results not only confirm the feasibility of the loop design, but also provide essential guidance and data support for future experimental implementation and the formulation of operational control strategies.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115522"},"PeriodicalIF":2.0,"publicationDate":"2025-11-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684567","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 : 2025-11-13DOI: 10.1016/j.fusengdes.2025.115513
Gwanghyeon Kwon , Yeong Woo Son , Jae-Uk Lee , Min Ho Chang , Jae Jung Urm , Jong Min Lee
In nuclear fusion fuel cycle, the isotope separation system (ISS) and water detritiation system (WDS) are critical for recycling hydrogen isotopologues but have rarely been analyzed in an integrated manner. To address this gap, we developed a steady-state simulation program that explicitly integrated ISS and WDS using an equation-oriented framework, incorporating rigorous thermodynamic, column design, and hydrodynamic models for cryogenic distillation (CD) columns and a liquid-phase catalytic exchange (LPCE) column. Extensive parametric studies on key variables such as feed flowrates, tank composition, and column feed stages were conducted to analyze their impacts on the separation performance of WDS and tritium inventory within ISS, highlighting the interdependent characteristics arising from the changes of the neighboring system. Furthermore, we identified the suitable design and operating variable sets, balancing high detritiation performance with low tritium inventory by systematically varying both the interfacial and non-interfacial variables. By offering a rigorous steady-state simulation that enables a comprehensive analysis of the integrated ISS–WDS process, the significance of this work lies in the simulator’s potential to analyze both processes from an integrated perspective. As a result, such findings can be utilized to define safe and efficient interface conditions during their integrated operations.
{"title":"Steady-state modeling and simulation of the integrated isotope separation and water detritiation systems: Parametric studies and performance analysis","authors":"Gwanghyeon Kwon , Yeong Woo Son , Jae-Uk Lee , Min Ho Chang , Jae Jung Urm , Jong Min Lee","doi":"10.1016/j.fusengdes.2025.115513","DOIUrl":"10.1016/j.fusengdes.2025.115513","url":null,"abstract":"<div><div>In nuclear fusion fuel cycle, the isotope separation system (ISS) and water detritiation system (WDS) are critical for recycling hydrogen isotopologues but have rarely been analyzed in an integrated manner. To address this gap, we developed a steady-state simulation program that explicitly integrated ISS and WDS using an equation-oriented framework, incorporating rigorous thermodynamic, column design, and hydrodynamic models for cryogenic distillation (CD) columns and a liquid-phase catalytic exchange (LPCE) column. Extensive parametric studies on key variables such as feed flowrates, tank composition, and column feed stages were conducted to analyze their impacts on the separation performance of WDS and tritium inventory within ISS, highlighting the interdependent characteristics arising from the changes of the neighboring system. Furthermore, we identified the suitable design and operating variable sets, balancing high detritiation performance with low tritium inventory by systematically varying both the interfacial and non-interfacial variables. By offering a rigorous steady-state simulation that enables a comprehensive analysis of the integrated ISS–WDS process, the significance of this work lies in the simulator’s potential to analyze both processes from an integrated perspective. As a result, such findings can be utilized to define safe and efficient interface conditions during their integrated operations.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"222 ","pages":"Article 115513"},"PeriodicalIF":2.0,"publicationDate":"2025-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684661","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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