Pub Date : 2025-02-12DOI: 10.1016/j.fusengdes.2025.114849
S. Ceccuzzi , DTT ICH Contributors, M. Aquilini , N. Badodi , A. Bonaventura , L. Boncagni , G. Camera , A. Cardinali , V. Casalegno , C. Castaldo , A. Cioffi , G. Di Gironimo , F. Di Paolo , M. Ferraris , V. Francalanza , D.L. Galindo-Huertas , G. Granucci , S. Greco , F. Grespan , F.G. Lanzotti , H.S. Wu
This paper describes the design and realisation status of the Ion Cyclotron Range of Frequency (ICRF) system of the Divertor Tokamak Test facility (DTT). DTT requires a large amount of additional heating partially provided by an ICRF system working in the frequency range from 60 to 90 MHz. The system will be modular, with each module aimed at coupling at least 3 MW for 50 s every hour to the DTT reference plasma as well as to contribute to wall cleaning tasks with lower power and higher duty cycle. Compared to existing and planned ICRF plants for tokamaks and stellarators, the radiofrequency system of DTT presents some peculiar features, mostly with reference to the technology of radiofrequency generators and to the mix of challenges the antenna design has to face like geometry decomposition, remote assembly and maintenance. The system design started some years ago and, in the last two years, 50+ collaborators contributed to make advances in its definition and development. Recently the ICRF system of DTT entered its realisation phase, with the issue of the first calls for tender, one of which for the procurement of two radiofrequency sources, while some aspects are still under design. This contribution gives a brief overview of the system architecture and focuses on the major advancements achieved in the latest years.
{"title":"Progress in the development of the ICRF system of DTT","authors":"S. Ceccuzzi , DTT ICH Contributors, M. Aquilini , N. Badodi , A. Bonaventura , L. Boncagni , G. Camera , A. Cardinali , V. Casalegno , C. Castaldo , A. Cioffi , G. Di Gironimo , F. Di Paolo , M. Ferraris , V. Francalanza , D.L. Galindo-Huertas , G. Granucci , S. Greco , F. Grespan , F.G. Lanzotti , H.S. Wu","doi":"10.1016/j.fusengdes.2025.114849","DOIUrl":"10.1016/j.fusengdes.2025.114849","url":null,"abstract":"<div><div>This paper describes the design and realisation status of the Ion Cyclotron Range of Frequency (ICRF) system of the Divertor Tokamak Test facility (DTT). DTT requires a large amount of additional heating partially provided by an ICRF system working in the frequency range from 60 to 90 MHz. The system will be modular, with each module aimed at coupling at least 3 MW for 50 s every hour to the DTT reference plasma as well as to contribute to wall cleaning tasks with lower power and higher duty cycle. Compared to existing and planned ICRF plants for tokamaks and stellarators, the radiofrequency system of DTT presents some peculiar features, mostly with reference to the technology of radiofrequency generators and to the mix of challenges the antenna design has to face like geometry decomposition, remote assembly and maintenance. The system design started some years ago and, in the last two years, 50+ collaborators contributed to make advances in its definition and development. Recently the ICRF system of DTT entered its realisation phase, with the issue of the first calls for tender, one of which for the procurement of two radiofrequency sources, while some aspects are still under design. This contribution gives a brief overview of the system architecture and focuses on the major advancements achieved in the latest years.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"213 ","pages":"Article 114849"},"PeriodicalIF":1.9,"publicationDate":"2025-02-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143395951","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-11DOI: 10.1016/j.fusengdes.2025.114852
Saira Gulfam, Kamran Ahmed, Asad Yaqoob Mian, Zahoor Ahmed
An in-depth analysis of buckling phenomena in a tokamak vacuum vessel of MT-II (Metallic Tokamak-II) is presented, offering critical insights into their implications for design and safety considerations. The study employs advanced modeling and simulation techniques to investigate the buckling behavior, providing a comprehensive understanding of structural responses under applied conditions. Findings from this analysis serve as a foundation for refining the vessel's design, enhancing safety protocols, and optimizing performance. The study underscores the importance of continuously improving design and operational strategies to ensure the robustness and reliability of the vacuum vessel.
The study explains non-destructive testing (NDT) procedures and inspection processes, highlighting areas that did not meet the strict ASME compliance requirements. Pressure testing results showed performance issues under different operating conditions, emphasizing the need for rigorous maintenance and inspection protocols.
{"title":"Analyzing buckling phenomena in an external pressure vessel: Implications for design and safety","authors":"Saira Gulfam, Kamran Ahmed, Asad Yaqoob Mian, Zahoor Ahmed","doi":"10.1016/j.fusengdes.2025.114852","DOIUrl":"10.1016/j.fusengdes.2025.114852","url":null,"abstract":"<div><div>An in-depth analysis of buckling phenomena in a tokamak vacuum vessel of MT-II (Metallic Tokamak-II) is presented, offering critical insights into their implications for design and safety considerations. The study employs advanced modeling and simulation techniques to investigate the buckling behavior, providing a comprehensive understanding of structural responses under applied conditions. Findings from this analysis serve as a foundation for refining the vessel's design, enhancing safety protocols, and optimizing performance. The study underscores the importance of continuously improving design and operational strategies to ensure the robustness and reliability of the vacuum vessel.</div><div>The study explains non-destructive testing (NDT) procedures and inspection processes, highlighting areas that did not meet the strict ASME compliance requirements. Pressure testing results showed performance issues under different operating conditions, emphasizing the need for rigorous maintenance and inspection protocols.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114852"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143388088","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-11DOI: 10.1016/j.fusengdes.2025.114867
Aleksandr Listopad, Yanis Andrebe, Patrick Blanchard, Stefano Coda, Frederic Dolizy, Jeremie Dubray, Basil P. Duval, Damien Fasel, Ambrogio Fasoli, Remy Jacquier, Alexander N. Karpushov, Yves Martin, Dmytry Mykytchuk, Marc Noel, Olivier Sauter, Ugo Siravo, Matthieu Toussaint, David Velasco De La Fuente, TCV team
The paper describes the operation and upgrade of the Heating Neutral Beam Injection (HNBI) system as well as the related infrastructure at the Tokamak Configuration Variable (TCV) during the last years.
In 2023, the installation of new TCV neutron shielding was completed, enabling the operation of two deuterium heating beams at their full power for 30 (instead of 5) high-performance TCV plasma shots.
The automation control capabilities of HNBIs have been enhanced through the integration of beam power scenario design into TCV discharge programming, along with real-time control of beams power based on real time plasma measurements of plasma during plasma discharge.
The beam ducts were re-designed for minimization of power loads impact. The integration of thermocouples was modified for better monitoring of the ducts heating.
Development of alternative ion optics for the first heating beam, designed for intermediate particle energies (up to ∼38 keV), has commenced. The first results of a beamlet geometry numerical optimization, based on beam formation IBSimu calculations, beam transportation through the duct, are considered respective the beam in-duct losses.
The prototype of spectral multichannel measurement setup for beam profile scanning and species composition evaluation was tested. The first images of the beam's spectral profile, providing fractional divergence consideration, have been analyzed.
{"title":"Present status of heating neutral beam injection system at TCV","authors":"Aleksandr Listopad, Yanis Andrebe, Patrick Blanchard, Stefano Coda, Frederic Dolizy, Jeremie Dubray, Basil P. Duval, Damien Fasel, Ambrogio Fasoli, Remy Jacquier, Alexander N. Karpushov, Yves Martin, Dmytry Mykytchuk, Marc Noel, Olivier Sauter, Ugo Siravo, Matthieu Toussaint, David Velasco De La Fuente, TCV team","doi":"10.1016/j.fusengdes.2025.114867","DOIUrl":"10.1016/j.fusengdes.2025.114867","url":null,"abstract":"<div><div>The paper describes the operation and upgrade of the Heating Neutral Beam Injection (HNBI) system as well as the related infrastructure at the Tokamak Configuration Variable (TCV) during the last years.</div><div>In 2023, the installation of new TCV neutron shielding was completed, enabling the operation of two deuterium heating beams at their full power for 30 (instead of 5) high-performance TCV plasma shots.</div><div>The automation control capabilities of HNBIs have been enhanced through the integration of beam power scenario design into TCV discharge programming, along with real-time control of beams power based on real time plasma measurements of plasma during plasma discharge.</div><div>The beam ducts were re-designed for minimization of power loads impact. The integration of thermocouples was modified for better monitoring of the ducts heating.</div><div>Development of alternative ion optics for the first heating beam, designed for intermediate particle energies (up to ∼38 keV), has commenced. The first results of a beamlet geometry numerical optimization, based on beam formation IBSimu calculations, beam transportation through the duct, are considered respective the beam in-duct losses.</div><div>The prototype of spectral multichannel measurement setup for beam profile scanning and species composition evaluation was tested. The first images of the beam's spectral profile, providing fractional divergence consideration, have been analyzed.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114867"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387852","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-11DOI: 10.1016/j.fusengdes.2025.114804
J.-H. You , H. Chae , R. Coppola , W. Gan , H. Greuner , M. Hofmann , S. Roccella , W. Woo
The stress state in a plasma-facing component (PFC) under high heat-flux (HHF) loads is the most important factor to determine the lifetime of the component. Stresses in a typical tungsten monoblock type PFC are produced by fabrication process and reactor operation where the component is subjected to HHF loads. In this study, both stress contributions were determined non-destructively by means of neutron diffraction technique. To this end, operational HHF loads were simulated using a high-power neutral hydrogen beam facility (GLADIS) to impose cyclic surface heating at 20 MW/m². A dedicated small-scale mock-up was fabricated applying hot radial pressing technique to join four tungsten blocks to a CuCrZr alloy cooling pipe via a soft 0.1 mm thick soft copper interlayer. This thin copper interlayer was used to simplify the residual stress profile for this preliminary test. The neutron diffraction measurements were carried out, at room temperature, at two different high-flux reactors: FRM II and the HANARO. Separate stress-relieved tungsten and CuCrZr samples were examined as reference state. The 3D stress tensor was determined in the same external block of the mock-up for both measurements, scanning it from the front face of the tungsten block towards the inner wall of the CuCrZr pipe. The results obtained at these two neutron sources are in good quantitative agreement. Comparing them with the stress profiles before thermal loading, it appears that after the HHF test at GLADIS compressive stresses up to -800 MPa developed in the tungsten block near the interlayer, while the CuCrZr pipe was scarcely affected, probably since the tungsten block accommodated most of the thermal impact. While stress measurements very close to the interlayer might have been affected by spatial resolution issues and error in the reference lattice parameter, the results of these experiments clearly indicate the significant impact of HHF loads on the stress profiles in the tungsten blocks.
{"title":"Impact of high heat flux loads on the residual stress in a tungsten-monoblock plasma-facing component","authors":"J.-H. You , H. Chae , R. Coppola , W. Gan , H. Greuner , M. Hofmann , S. Roccella , W. Woo","doi":"10.1016/j.fusengdes.2025.114804","DOIUrl":"10.1016/j.fusengdes.2025.114804","url":null,"abstract":"<div><div>The stress state in a plasma-facing component (PFC) under high heat-flux (HHF) loads is the most important factor to determine the lifetime of the component. Stresses in a typical tungsten monoblock type PFC are produced by fabrication process and reactor operation where the component is subjected to HHF loads. In this study, both stress contributions were determined non-destructively by means of neutron diffraction technique. To this end, operational HHF loads were simulated using a high-power neutral hydrogen beam facility (GLADIS) to impose cyclic surface heating at 20 MW/m². A dedicated small-scale mock-up was fabricated applying hot radial pressing technique to join four tungsten blocks to a CuCrZr alloy cooling pipe via a soft 0.1 mm thick soft copper interlayer. This thin copper interlayer was used to simplify the residual stress profile for this preliminary test. The neutron diffraction measurements were carried out, at room temperature, at two different high-flux reactors: FRM II and the HANARO. Separate stress-relieved tungsten and CuCrZr samples were examined as reference state. The 3D stress tensor was determined in the same external block of the mock-up for both measurements, scanning it from the front face of the tungsten block towards the inner wall of the CuCrZr pipe. The results obtained at these two neutron sources are in good quantitative agreement. Comparing them with the stress profiles before thermal loading, it appears that after the HHF test at GLADIS compressive stresses up to -800 MPa developed in the tungsten block near the interlayer, while the CuCrZr pipe was scarcely affected, probably since the tungsten block accommodated most of the thermal impact. While stress measurements very close to the interlayer might have been affected by spatial resolution issues and error in the reference lattice parameter, the results of these experiments clearly indicate the significant impact of HHF loads on the stress profiles in the tungsten blocks.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114804"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143378593","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-11DOI: 10.1016/j.fusengdes.2025.114861
J. Boscary , P. Vinoni , D. Marzullo , A. Quartararo , A. Cufar , S. Renard , M. Kannamüller , S. Wiesen , G. Aiello , C. Bachmann , D. Leichtle , P. Gallina , R. Neu
The goal of the European plasma Volumetric Neutron Source (VNS), which is a 14 MeV n-source, is to test and validate technological solutions of breeding blankets in an environment representative of a future fusion power plant such as DEMO. One of the assessed magnetic configurations is a tokamak device with a radius of ≈ 2.5 m that produces a D-T fusion power of ≈ 30 MW. The architecture of VNS integrates a single null divertor located at the bottom of the vacuum vessel. The ITER-like divertor is designed to be actively water-cooled for stationary operation and is made up with 36 cassette modules. Each module has a cassette body with a set of plasma facing components (PFCs) with tungsten as plasma facing material. The PFC arrangement is a dome positioned in the central part with reflector plates on both sides located between inboard and outboard target striking surfaces. The selected PFC technological solutions are: the plasma facing units (PFUs) of the targets are armored with tungsten monoblocs bonded onto a cooling pipe made of CuCrZr with an inserted twisted tape, the PFUs of the dome are made up with tungsten flat tiles bonded onto a CuCrZr hypervapotron cooling structure. This paper presents the status of the development of the divertor conceptual design. Preliminary analyses confirmed that the design is compatible with plasma scenarios foreseen for VNS operation.
{"title":"Divertor conceptual design of the European Volumetric Neutron Source","authors":"J. Boscary , P. Vinoni , D. Marzullo , A. Quartararo , A. Cufar , S. Renard , M. Kannamüller , S. Wiesen , G. Aiello , C. Bachmann , D. Leichtle , P. Gallina , R. Neu","doi":"10.1016/j.fusengdes.2025.114861","DOIUrl":"10.1016/j.fusengdes.2025.114861","url":null,"abstract":"<div><div>The goal of the European plasma Volumetric Neutron Source (VNS), which is a 14 MeV n-source, is to test and validate technological solutions of breeding blankets in an environment representative of a future fusion power plant such as DEMO. One of the assessed magnetic configurations is a tokamak device with a radius of ≈ 2.5 m that produces a D-T fusion power of ≈ 30 MW. The architecture of VNS integrates a single null divertor located at the bottom of the vacuum vessel. The ITER-like divertor is designed to be actively water-cooled for stationary operation and is made up with 36 cassette modules. Each module has a cassette body with a set of plasma facing components (PFCs) with tungsten as plasma facing material. The PFC arrangement is a dome positioned in the central part with reflector plates on both sides located between inboard and outboard target striking surfaces. The selected PFC technological solutions are: the plasma facing units (PFUs) of the targets are armored with tungsten monoblocs bonded onto a cooling pipe made of CuCrZr with an inserted twisted tape, the PFUs of the dome are made up with tungsten flat tiles bonded onto a CuCrZr hypervapotron cooling structure. This paper presents the status of the development of the divertor conceptual design. Preliminary analyses confirmed that the design is compatible with plasma scenarios foreseen for VNS operation.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114861"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387853","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}
A new dry process using microwaves was proposed for refining beryllium from poorly soluble beryl, resulting in the complete dissolution of beryl at low temperatures and under atmospheric pressure. The microwave heating source resulted in a significantly higher dissolved fraction of beryllium than when an external heating source was used. The apparent activation energy for beryllium leaching from beryl was approximately 30 % lower for microwave heating than for external heating. The reaction acceleration effect observed during microwave heating can be attributed to the formation of localized high-temperature spots owing to microwave irradiation. The influence of non-thermal effects was considered negligible in the proposed system.
{"title":"Development of new dry dissolution process of poorly soluble beryl and its acceleration effect by microwave heating","authors":"Suguru Nakano , Jae-Hwan Kim , Yoshiaki Akatsu , Shota Yokohama , Taehyun Hwang , Yutaka Sugimoto , Ryuta Kasada , Masaru Nakamichi","doi":"10.1016/j.fusengdes.2025.114860","DOIUrl":"10.1016/j.fusengdes.2025.114860","url":null,"abstract":"<div><div>A new dry process using microwaves was proposed for refining beryllium from poorly soluble beryl, resulting in the complete dissolution of beryl at low temperatures and under atmospheric pressure. The microwave heating source resulted in a significantly higher dissolved fraction of beryllium than when an external heating source was used. The apparent activation energy for beryllium leaching from beryl was approximately 30 % lower for microwave heating than for external heating. The reaction acceleration effect observed during microwave heating can be attributed to the formation of localized high-temperature spots owing to microwave irradiation. The influence of non-thermal effects was considered negligible in the proposed system.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114860"},"PeriodicalIF":1.9,"publicationDate":"2025-02-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143377891","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 : 2025-02-09DOI: 10.1016/j.fusengdes.2025.114846
L. Bühler, C. Courtessole, C. Koehly, B. Lyu, C. Mistrangelo
The distribution of electric potential on the surface of a scaled mock-up of the water-cooled lead lithium test blanket module for ITER has been measured in magnetohydrodynamic experiments using the MEKKA facility at the Karlsruhe Institute of Technology (KIT). The electric potential is a physical quantity that can be recorded with high accuracy. For strong magnetic fields, i.e. for high Hartmann numbers, the potential may be considered as approximate hydrodynamic stream-function of the liquid metal flow, from which the core velocity is obtained by differentiation. Moreover, experimental potential data may be directly compared with numerical simulations for both validation of computational tools and interpretation of measurements. Experimental and numerical data of potential confirm anticipations from measured pressure values that the major fraction of the flow in breeder units occurs near both ends of the module, while in a larger central domain the flow is almost negligible.
{"title":"Electric potential on a WCLL TBM mock-up in MHD experiments as indication for flow distribution in breeder units","authors":"L. Bühler, C. Courtessole, C. Koehly, B. Lyu, C. Mistrangelo","doi":"10.1016/j.fusengdes.2025.114846","DOIUrl":"10.1016/j.fusengdes.2025.114846","url":null,"abstract":"<div><div>The distribution of electric potential on the surface of a scaled mock-up of the water-cooled lead lithium test blanket module for ITER has been measured in magnetohydrodynamic experiments using the MEKKA facility at the Karlsruhe Institute of Technology (KIT). The electric potential is a physical quantity that can be recorded with high accuracy. For strong magnetic fields, i.e. for high Hartmann numbers, the potential may be considered as approximate hydrodynamic stream-function of the liquid metal flow, from which the core velocity is obtained by differentiation. Moreover, experimental potential data may be directly compared with numerical simulations for both validation of computational tools and interpretation of measurements. Experimental and numerical data of potential confirm anticipations from measured pressure values that the major fraction of the flow in breeder units occurs near both ends of the module, while in a larger central domain the flow is almost negligible.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114846"},"PeriodicalIF":1.9,"publicationDate":"2025-02-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143373110","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 International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source (IFMIF-DONES) is an advanced linear accelerator designed to produce high-intensity neutron fluxes for analyzing materials under conditions similar to the DEMO fusion reactor. It accelerates a deuteron beam onto a liquid lithium target, generating a neutron flux with a peak energy of 14 MeV, which then interacts with specimens in the High Flux Test Module (HFTM). Due to the significant neutron activation, the Test Cell necessitates remote handling for maintenance, requiring custom-designed components. The maintenance procedures for these components must be developed, tested, and analyzed ex-novo. The HFTM electrical connectors bridge, previously validated, can be employed as a helpful test benchmark for studying and improving the maintenance procedure development and simulation methodology.
This paper details the development, simulation, and verification of a remote handling maintenance procedure for the HFTM electrical connectors bridge. The procedure was developed using functional analysis and simulated with Petri nets in the Modelica simulation language, offering a clear visualization of each step and valuable metrics like total maintenance time and time required per operation. Simulations highlighted optimization opportunities, particularly in alignment procedures. The experimental activities were carried out at the Divertor Refurbishment Platform (DRP) laboratory of ENEA C.R. Brasimone. These confirmed the predicted criticalities from the functional analysis and aligned with the Petri net simulation results. The experimental analysis yielded positive results, where the possible optimizations to the alignment system found by the Petri nets were tested and verified.
{"title":"Development and modeling of a remote handling maintenance procedure for the IFMIF-DONES high flux test module electrical connectors bridge","authors":"Gabriele Benzoni , Carolina Introini , Gioacchino Miccichè , Claudio Tripodo , Antonio Cammi","doi":"10.1016/j.fusengdes.2025.114857","DOIUrl":"10.1016/j.fusengdes.2025.114857","url":null,"abstract":"<div><div>The International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source (IFMIF-DONES) is an advanced linear accelerator designed to produce high-intensity neutron fluxes for analyzing materials under conditions similar to the DEMO fusion reactor. It accelerates a deuteron beam onto a liquid lithium target, generating a neutron flux with a peak energy of 14 MeV, which then interacts with specimens in the High Flux Test Module (HFTM). Due to the significant neutron activation, the Test Cell necessitates remote handling for maintenance, requiring custom-designed components. The maintenance procedures for these components must be developed, tested, and analyzed <em>ex-novo</em>. The HFTM electrical connectors bridge, previously validated, can be employed as a helpful test benchmark for studying and improving the maintenance procedure development and simulation methodology.</div><div>This paper details the development, simulation, and verification of a remote handling maintenance procedure for the HFTM electrical connectors bridge. The procedure was developed using functional analysis and simulated with Petri nets in the Modelica simulation language, offering a clear visualization of each step and valuable metrics like total maintenance time and time required per operation. Simulations highlighted optimization opportunities, particularly in alignment procedures. The experimental activities were carried out at the Divertor Refurbishment Platform (DRP) laboratory of ENEA C.R. Brasimone. These confirmed the predicted criticalities from the functional analysis and aligned with the Petri net simulation results. The experimental analysis yielded positive results, where the possible optimizations to the alignment system found by the Petri nets were tested and verified.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114857"},"PeriodicalIF":1.9,"publicationDate":"2025-02-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143373109","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-08DOI: 10.1016/j.fusengdes.2025.114840
A. Adriaens , F. Durodié , V. Maquet , S. Deshpande , D. López-Rodríguez , M. Verstraeten , A. Goriaev , K. Crombé , J. Buermans , S. Brezinsek
The TOMAS device is equipped with an Ion Cyclotron Range-of-Frequency (ICRF) system enabling systematic investigations of Ion Cyclotron Wall Conditioning (ICWC) in a toroidal geometry complementing plasma-wall interaction and plasma production research in larger devices.
The ICRF system on the TOMAS device has a varying load as the plasma may change in density, pressure, magnetic field and species (H/D/He/Ar) depending on the required experimental conditions. Therefore, a T-matching capacitor network is installed. To correctly control these variable capacitors in order to achieve efficient power transmission to the plasma, in this paper, multiple algorithms were developed. The first kind of algorithm requires voltage probes along a coaxial line connecting the power source to the antenna, for these kinds algorithms at least 4 inputs are necessary: at least 3 voltage probe measurements along the line and the amplitude of the forward voltage, either measured directly or inferred using the requested power gain. For the second form of algorithm the amplitude of both the forward and reflected wave need to be measured as well as their phase difference. System parameters were investigated over the course of simulations and afterwards tests of one of the algorithms were carried out on the TOMAS machine, confirming its effectiveness.
{"title":"An automatic matching system for the ICRF antenna at TOMAS: Development and experimental proof","authors":"A. Adriaens , F. Durodié , V. Maquet , S. Deshpande , D. López-Rodríguez , M. Verstraeten , A. Goriaev , K. Crombé , J. Buermans , S. Brezinsek","doi":"10.1016/j.fusengdes.2025.114840","DOIUrl":"10.1016/j.fusengdes.2025.114840","url":null,"abstract":"<div><div>The TOMAS device is equipped with an Ion Cyclotron Range-of-Frequency (ICRF) system enabling systematic investigations of Ion Cyclotron Wall Conditioning (ICWC) in a toroidal geometry complementing plasma-wall interaction and plasma production research in larger devices.</div><div>The ICRF system on the TOMAS device has a varying load as the plasma may change in density, pressure, magnetic field and species (H/D/He/Ar) depending on the required experimental conditions. Therefore, a T-matching capacitor network is installed. To correctly control these variable capacitors in order to achieve efficient power transmission to the plasma, in this paper, multiple algorithms were developed. The first kind of algorithm requires voltage probes along a coaxial line connecting the power source to the antenna, for these kinds algorithms at least 4 inputs are necessary: at least 3 voltage probe measurements along the line and the amplitude of the forward voltage, either measured directly or inferred using the requested power gain. For the second form of algorithm the amplitude of both the forward and reflected wave need to be measured as well as their phase difference. System parameters were investigated over the course of simulations and afterwards tests of one of the algorithms were carried out on the TOMAS machine, confirming its effectiveness.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114840"},"PeriodicalIF":1.9,"publicationDate":"2025-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143350738","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-02-08DOI: 10.1016/j.fusengdes.2025.114863
Mingyuan Yang , Quan Zhou , Hongbin Huang , Jie Liu , Hongtao Pan , Yong Cheng , Zongkuan Kang , Zhongxu Hu , Youmin Hu
The CFETR Multi-Purpose Overload Robot (CMOR), a 7- DOF manipulator with 7 joints, serves as a crucial component in the China Fusion Engineering Test Reactor (CFETR). Each joint integrates electrical and mechanical components seamlessly including sensors, motors, and planetary reducers. The robot is characterized by its heavy-load capacity, high precision, and compact structure, placing significant demands on both the bearings and the gears of the reducers. Therefore, fault prediction for bearings of CMOR is essential and beneficial for normal operation of the whole machinery. This paper presents a novel fault prediction method for CMOR bearing based on parameter-optimized variational mode decomposition (VMD) and autocorrelation function (ACF). Firstly, a new fitness function is defined based on ACF. Then, optimize parameters of VMD by slime mold algorithm (SMA). Next, VMD is performed on the signal and suitable modes are selected for signal reconstruction. Then Hilbert transform (HT) is employed to the reconstructed signal for fault prediction. Finally, in order to evaluate the effectiveness of each envelope spectrum, a quantitative indicator (Q) is defined to measure the proximity between the theoretical fault characteristic frequency (or its 2nd to 4th harmonics) and the center frequencies computed in the spectrum. Comparative experiments through the quantitative indicator Q demonstrate that the superiority of the proposed method has validated over two other methods. It is also demonstrated that the proposed method significantly benefits CMOR bearings' future fault prediction.
{"title":"A fault prediction method for CMOR bearings based on parameter-optimized variational mode decomposition and autocorrelation function","authors":"Mingyuan Yang , Quan Zhou , Hongbin Huang , Jie Liu , Hongtao Pan , Yong Cheng , Zongkuan Kang , Zhongxu Hu , Youmin Hu","doi":"10.1016/j.fusengdes.2025.114863","DOIUrl":"10.1016/j.fusengdes.2025.114863","url":null,"abstract":"<div><div>The CFETR Multi-Purpose Overload Robot (CMOR), a 7- DOF manipulator with 7 joints, serves as a crucial component in the China Fusion Engineering Test Reactor (CFETR). Each joint integrates electrical and mechanical components seamlessly including sensors, motors, and planetary reducers. The robot is characterized by its heavy-load capacity, high precision, and compact structure, placing significant demands on both the bearings and the gears of the reducers. Therefore, fault prediction for bearings of CMOR is essential and beneficial for normal operation of the whole machinery. This paper presents a novel fault prediction method for CMOR bearing based on parameter-optimized variational mode decomposition (VMD) and autocorrelation function (ACF). Firstly, a new fitness function is defined based on ACF. Then, optimize parameters of VMD by slime mold algorithm (SMA). Next, VMD is performed on the signal and suitable modes are selected for signal reconstruction. Then Hilbert transform (HT) is employed to the reconstructed signal for fault prediction. Finally, in order to evaluate the effectiveness of each envelope spectrum, a quantitative indicator (<em>Q</em>) is defined to measure the proximity between the theoretical fault characteristic frequency (or its 2nd to 4th harmonics) and the center frequencies computed in the spectrum. Comparative experiments through the quantitative indicator <em>Q</em> demonstrate that the superiority of the proposed method has validated over two other methods. It is also demonstrated that the proposed method significantly benefits CMOR bearings' future fault prediction.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"212 ","pages":"Article 114863"},"PeriodicalIF":1.9,"publicationDate":"2025-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143373106","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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