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Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors

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A special issue of Applied Sciences (ISSN 2076-3417). This special issue belongs to the section "Materials Science and Engineering".

Deadline for manuscript submissions: closed (20 June 2022) | Viewed by 13724

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Dr. Pietro ARENA

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Guest Editor

Italian National Agency for New Technologies, Energy and Sustainable Economic Development, ENEA FSN-ING-SIS, C.R. Brasimone, 40032 Camugnano, BO, Italy
Interests: nuclear fusion reactors; thermo-mechanics; finite element modelling; nuclear energy

Prof. Dr. Pietro Alessandro Di Maio

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Guest Editor

Department of Engineering, University of Palermo, 90128 Palermo, Italy
Interests: nuclear fusion reactors; thermo-mechanics; pipe stress analysis; thermofluid-dynamics; neutronics; finite element modelling; finite volume modelling; nuclear power plants

Special Issue Information

Dear Colleagues,

Nuclear fusion is one of the most attractive technologies for the achievement of electricity production in a safe, sustainable and renewable way, while maintaining a carbon-free approach. On the other hand, the design of a nuclear fusion reactor represents a very challenging activity due to the very demanding operational conditions and the number of cutting-edge technologies to be employed in these kind of reactors. Indeed, a nuclear reactor is a device where the most extreme conditions are reached, both in terms of neutronic, thermal, magnetic and mechanical points of view. In few meters, the highest and lowest achievable temperatures are experienced, from hundreds of millions of °K in the plasma to a few °K in the supercritical helium cooling the magnets. For all these reasons, the proper design of the different systems constituting a nuclear fusion reactor is fundamental for the achievement of this challenging goal.

The scope of the present Special Issue is, therefore, to collect submissions reporting the state of the art of R&D activity on nuclear fusion reactors, investigating the main issues related to the structural and thermo-mechanical design, as well as to look for advanced and innovative methods for the design of reactor components.

Dr. Pietro ARENA
Prof. Dr. Pietro Alessandro Di Maio
Guest Editors

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Keywords

nuclear fusion reactors
structural analysis
pipe stress analysis
thermo-mechanics
finite element method
nuclear design

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Published Papers (6 papers)

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Editorial

Jump to: Research

2 pages, 170 KiB

Open AccessEditorial

Special Issue on Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors

by Pietro Arena and Pietro Alessandro Di Maio

Appl. Sci. 2022, 12(24), 12562; https://doi.org/10.3390/app122412562 - 8 Dec 2022

Cited by 2 | Viewed by 1113

Abstract

Nuclear fusion is one of the most promising technologies to be adopted for the production of electricity [...] Full article

(This article belongs to the Special Issue Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors)

Research

Jump to: Editorial

20 pages, 10534 KiB

Open AccessArticle

A Combined Electromagnetic and Mechanical Approach for EU-DEMO Toroidal Field Coils

by Lorenzo Giannini, Daniela P. Boso and Valentina Corato

Appl. Sci. 2022, 12(6), 2766; https://doi.org/10.3390/app12062766 - 8 Mar 2022

Cited by 6 | Viewed by 3042

Abstract

The roadmap to fusion electricity of the European scientific program aims at the realization of the future DEMOnstration (DEMO) fusion power plant. In 2020, the pre-concept design phase of DEMO was completed, defining the concept and characteristics of the main magnets and structures of the machine. Sixteen toroidal D-shaped magnets, six poloidal annular coils and a central solenoid constitute the functioning system core. The reactor is subjected to huge mechanical loads, mainly due to the Lorentz force produced by the combination of the high magnetic fields and operative currents. As a consequence, the loading conditions are extremely demanding for the structural components, and it is crucial to complete a comprehensive static and fatigue assessment before proceeding with the next design iteration. This work focuses on the electromagnetic and structural analyses performed on the toroidal field coil system and its support structures to present the methodological approach developed. Exploiting the finite element method, a three-dimensional model has been defined to obtain the electromagnetic loads on the main time points of the reference plasma scenario and then transfer them to a related 3D structural model, corresponding to the discretization of the electromagnetic one. The structural model was used to obtain the displacement and stress fields at the various time points to perform the mechanical evaluation as well as the fatigue assessment. Full article

(This article belongs to the Special Issue Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors)

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23 pages, 15936 KiB

Open AccessArticle

Thermo-Mechanical Analysis and Design Update of the Top Cap Region of the DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment

by Gaetano Bongiovì, Salvatore Giambrone, Ilenia Catanzaro, Pietro Alessandro Di Maio and Pietro Arena

Appl. Sci. 2022, 12(3), 1564; https://doi.org/10.3390/app12031564 - 31 Jan 2022

Cited by 3 | Viewed by 1949

Abstract

Within the framework of the EUROfusion research and development activities, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) is one of the two candidates to be chosen as the driver blanket for the European DEMO nuclear fusion reactor. Hence, an intense research work is currently ongoing throughout the EU to develop a robust conceptual design able to fulfil the design requirements selected at the end of the DEMO pre-conceptual design phase. In this work, the thermo-mechanical analysis and the design update of the top cap (TC) region of the DEMO WCLL Central Out-board Blanket (COB) segment is presented. The scope of the work is to find a design solution of the WCLL COB TC region able to fulfil the design requirements, prescribed by the reference RCC-MRx code, under the selected nominal and accidental steady state loading scenarios. The activity herein presented moved from the WCLL COB reference design, purposely modified in compliance with the adopted thermal and mechanical requirements in order to attain a robust TC region geometric layout. In the end, a geometric configuration called “TC region-mod++” was determined, foreseeing a TC able to safely withstand both nominal and accidental loads. Nevertheless, some criticalities still hold in the internal stiffening plates and, therefore, further and finer analysis are necessary to fully match the goal. In any case, it was also found that the proposed approach for the design update is promising and worthy to be further pursued. The work was performed following a theoretical–numerical approach based on the finite element method (FEM) and adopting the quoted Ansys commercial FEM code. Full article

(This article belongs to the Special Issue Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors)

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20 pages, 9450 KiB

Open AccessFeature PaperArticle

Analysis of the Thermo-Mechanical Behaviour of the EU DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment under an Optimized Thermal Field

by Ilenia Catanzaro, Gaetano Bongiovì and Pietro Alessandro Di Maio

Appl. Sci. 2022, 12(3), 1356; https://doi.org/10.3390/app12031356 - 27 Jan 2022

Cited by 13 | Viewed by 1601

Abstract

Within the framework of the EUROfusion research activities on the DEMO Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) design, a research study was performed to preliminarily optimize, from the thermal point of view, the WCLL Central Outboard Blanket (COB) segment in order to investigate its structural behaviour under a realistic thermal field. In particular, a study of thermal analyses was performed to optimize the Double Walled Tubes and Segment Box cooling channels’ geometric configurations along the poloidal extension of the WCLL COB segment, in order to obtain a spatial temperature distribution fulfilling the thermal design requirement. Then, the thermo-mechanical analysis of the WCLL COB segment under Normal Operation (NO, representing nominal conditions), Upper Vertical Displacements Event (UVDE, representing a plasma disruption event) and Over-Pressurization (OP, representing an in-box loss of coolant accident) scenarios were carried out, assuming the previously obtained thermal field, to realistically predict displacement and stress fields. Finally, a stress linearization procedure allowed comparing the stress values obtained in some critical regions of the structure with the criteria prescribed by the reference design standard RCC-MRx. A theoretical–numerical approach based on the Finite Element Method (FEM) was followed using the commercial code Abaqus v. 6.14. Full article

(This article belongs to the Special Issue Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors)

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14 pages, 6325 KiB

Open AccessArticle

Analysis of the New Thomson Scattering Diagnostic System on WEST Tokamak

by Brahim Chelihi, Gilles Colledani, Louis Doceul, Nicolas Lefèvre, Tristan Batal, Silvia Garitta, Frédéric Faïsse, Jean-Marc Verger and Antony Bec

Appl. Sci. 2022, 12(3), 1318; https://doi.org/10.3390/app12031318 - 26 Jan 2022

Cited by 2 | Viewed by 2838

Abstract

The French tokamak WEST supports the ITER design and operation. IRFM is designing a new Thomson scattering diagnostic to measure plasma density and temperature profiles. The diagnostic system consists of an endoscope inside a vacuum vessel, composed of actively cooled optical components. In order to validate and guarantee the diagnostic performances during normal operations, mechanical, thermal, hydraulic and vibratory behavior must be checked. Moreover, perpendicular displacement of the optical surface shall not be higher than 40 µm. Since this diagnostic operates in near infrared light, the temperature of all components must stay lower than 200 °C as not to bias the measurements. The differences in water temperature and pressure between the inlet and outlet of the diagnostic must be lower than 50 °C and 5.6 bar, respectively. The natural frequencies of the structure must be higher than 20 Hz and far enough from the frequency of external components. In this study, the worst radiative plasma scenario was chosen. The results of this study validate the accuracy of the measurements. Before manufacturing, electromagnetic disruption events must also be considered. Full article

(This article belongs to the Special Issue Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors)

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15 pages, 7334 KiB

Open AccessArticle

Application of Inelastic Method and Its Comparison with Elastic Method for the Assessment of In-Box LOCA Event on EU DEMO HCPB Breeding Blanket Cap Region

by Anoop Retheesh, Francisco A. Hernández and Guangming Zhou

Appl. Sci. 2021, 11(19), 9104; https://doi.org/10.3390/app11199104 - 30 Sep 2021

Cited by 4 | Viewed by 1916

Abstract

The Helium Cooled Pebble Bed (HCPB) breeding blanket, being developed by the Karlsruhe Institute of Technology (KIT) and its partners is one of the two driver blanket candidates to be selected for the European demonstration fusion power plant (EU DEMO). The in-box Loss of Coolant Accident (LOCA) is a postulated initiating event of the breeding blanket (BB) that must be accounted within the design basis. In this paper, the BB cap region is analyzed for its ability to withstand an in-box LOCA event. Initially, an assessment is performed using conventional elastic design codes for nuclear pressure vessels. However, it is thought that the elastic rules are not ‘equipped’ to assess the material damage modes which are essentially inelastic. Therefore, a non-linear inelastic analysis is further performed to better understand the damage in the material. Two predominant inelastic failure modes are thought to be relevant and addressed: exhaustion of ductility and plastic flow localization. While the design of HCPB BB has been predominantly based on the elastic design-by-analysis studies, results from the present study show that the elastic rules may be overly conservative for the given material and loading and could lead to inefficient designs. To our knowledge, this study is the first attempt to investigate the structural integrity of the European DEMO blankets under in-box LOCA conditions using the inelastic methods. Full article

(This article belongs to the Special Issue Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors)

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