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Advances in Fusion Engineering and Design

A special issue of Applied Sciences (ISSN 2076-3417). This special issue belongs to the section "Applied Physics General".

Deadline for manuscript submissions: closed (10 August 2023) | Viewed by 33624

Special Issue Editor

Dr. Arkady Serikov

E-Mail Website
Guest Editor
Karlsruhe Institute of Technology (KIT)
Interests: Monte Carlo method; Monte Carlo variance reduction techniques; Radiation transport; Fusion neutronics; Radiation damage; Radiation shielding analyses; Nuclear analyses; Activation calculations; Nuclear fusion facilities ITER, IFMIF-DONES, DEMO; Neutronics analyses; Deuteron accelerators; Deuteron beam; Liquid lithium target; Nuclear heating; Neutron flux.
Special Issues, Collections and Topics in MDPI journals

Special Issue Information

Dear colleagues,

We believe that this century activities in the thermonuclear fusion field are going to reach an unprecedented level of engineering and technological achievements. Scientific fundamental hypotheses and theories now have a chance to be realized not only in computational models but also in real fusion devices and facilities. Advances in design solutions are implemented in the construction of large-size ITER tokamak. This is a very exciting time when fusion dreams come true. Normally, to speed-up the R&D processes we participated in face-to-face fusion conferences. However, now we are limited by online communications, and the need for broader and open communications is evident. In order to present your achievements in computational and experimental fusion science and technology, we at the MDPI Open Access journal "Applied Sciences" decided to organize a Special Issue with the rather broad title “Advances in Fusion Engineering and Design”. Writing your paper, please keep in mind that a particular type of thermonuclear system does not limit the scope of this Special Issue. It could be a tokamak (e.g. ITER, DEMO, JET, ARIES-ST, JT-60SA), stellarator (Helias, W7X), laser-based facility NIF, or an accelerator-driven system like IFMIF-DONES. Important is to present the innovative results which are advancing the development of fusion facilities. These results being compiled in one Special Issue and wider disseminated with open-access will be targeted on fusion specialists, readers MDPI journals, and qualified audience. By showing the progress in solving actual engineering tasks of fusion devices design, we are going to promote fusion as a realistic source of energy.

Topics for this Special Issue include, but are not limited to the following:

  • Applied science of magnetic and inertial fusion energy;
  • Advances in fusion technology and engineering;
  • Design studies for fusion experiments and devices;
  • Progress in blanket designs (shielding and breeding);
  • Plasma facing components and use of beryllium;
  • Plasma material interactions;
  • Liquid metals in blanket design;
  • Fusion fuel cycle and tritium technology;
  • Plasma heating and current drive systems;
  • Plasma enabling technology;
  • Development of plasma diagnostics;
  • Test facilities for materials and components;
  • Materials and nuclear technologies;
  • Fusion neutronics;
  • Thermal hydraulics for fusion components;
  • System analysis and model integration;
  • Magnets in fusion devices;
  • Advances in fusion safety;
  • Waste management for fusion facilities;
  • Fusion enterprise / private fusion companies / new concepts.

Dr. Arkady Serikov
Guest Editor

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Applied Sciences is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 2400 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • Fusion technology
  • Fusion engineering and design
  • Neutronics
  • Thermal hydraulics
  • System analysis
  • Computer modelling
  • Model integration
  • Blankets
  • Liquid-metal blankets
  • MHD thermofluids
  • Breeding blankets
  • Materials
  • Magnets
  • Fusion fuel cycle
  • Tritium
  • Plasma facing components
  • Beryllium
  • Plasma material interactions
  • Plasma
  • Plasma heating
  • ITER
  • DEMO
  • JET
  • ARIES-ST
  • JT-60SA
  • Helias
  • W7X
  • NIF
  • IFMIF-DONES
  • Diagnostics
  • Fusion safety
  • Waste management
  • Private fusion

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Published Papers (16 papers)

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23 pages, 17158 KiB  
Article
Exploratory Thermo-Mechanical Assessment of the Bottom Cap Region of the EU DEMO Water-Cooled Lead Lithium Central Outboard Blanket Segment
by Gaetano Bongiovì, Ilenia Catanzaro, Pietro Alessandro Di Maio, Salvatore Giambrone, Alberto Gioè, Pietro Arena and Lorenzo Melchiorri
Appl. Sci. 2023, 13(17), 9812; https://doi.org/10.3390/app13179812 - 30 Aug 2023
Viewed by 766
Abstract
The Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB) is one of the two BB concept candidates to be selected as the driver blanket for the EU DEMO fusion reactor. In this regard, the development of a sound architecture of the WCLL Central Outboard [...] Read more.
The Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB) is one of the two BB concept candidates to be selected as the driver blanket for the EU DEMO fusion reactor. In this regard, the development of a sound architecture of the WCLL Central Outboard Blanket (COB) Segment, ensuring the fulfilment of the thermal and structural design requirements, is one of the main goals of the EUROfusion consortium. To this purpose, an exploratory research campaign has been launched to preliminarily investigate the thermo-mechanical performances of the Bottom Cap (BC) region of the WCLL COB segment because of its peculiarities making its design different from the other regions. The assessment has been carried out considering the nominal BB operating conditions, the Normal Operation (NO) scenario, as well as a steady-state scenario derived from the in-box LOCA accident, the Over-Pressurization (OP) scenario. Starting from the reference geometric layout of the WCLL COB BC region, a first set of analyses has been launched in order to evaluate its structural performances under a previously calculated thermal field and to select potential geometric improvements. Then, the analysis of a complete BC region was conducted from both the thermal and structural standpoints, evaluating its structural behaviour in compliance with the RCC-MRx code. Finally, after some iterations and geometric updates, a promising geometric layout of the BC region has been obtained even though some criticalities still persist in the internal Stiffening Plates and First Wall. However, the obtained results clearly showed that the proposed layout is worthy to be further assessed to achieve a robust enough configuration. The work has been 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 Advances in Fusion Engineering and Design)
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22 pages, 7823 KiB  
Article
Calculation of Consistent Plasma Parameters for DEMO-FNS Using Ionic Transport Equations and Simulation of the Tritium Fuel Cycle
by Sergey Ananyev and Andrei Kukushkin
Appl. Sci. 2023, 13(14), 8552; https://doi.org/10.3390/app13148552 - 24 Jul 2023
Cited by 1 | Viewed by 1269
Abstract
Modeling the D and T fluxes in Fusion Neutron Source based on a tokamak fuel cycle systems was performed consistently with the core and divertor plasma. An indirect integration of ASTRA, SOLPS4.3, and FC-FNS codes is used. The feedback coupling is realized between [...] Read more.
Modeling the D and T fluxes in Fusion Neutron Source based on a tokamak fuel cycle systems was performed consistently with the core and divertor plasma. An indirect integration of ASTRA, SOLPS4.3, and FC-FNS codes is used. The feedback coupling is realized between the pumping and puffing systems in the form of changes in the isotopic composition of the core and edge plasma. In the ASTRA code, instead of electrons, ions were used in the particle transport equations. This allows better estimates of the flows of the D/T components of the fuel that have to be provided by the gas puffing and processing systems. The particle flows into the plasma from pellets, required to maintain the target plasma density <ne> = (6–8) × 1019 m−3 are 1022 particles/s. In the majority of the working range of parameters, additional ELM stimulation is necessary (by ~1-mm3-size pellets from the low magnetic field side) in order to maintain the controlled energy losses at the level δWELM~0.5 MJ. For the starting load of the FC and steady-state operation of the facility, up to 500 g of tritium are required taking into account the radioactive decay losses. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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25 pages, 4768 KiB  
Article
Activation Analyses of Disposal Options for Irradiated Be12Ti
by Pavel Pereslavtsev, Pierre Cortes and Joelle Elbez-Uzan
Appl. Sci. 2023, 13(13), 7534; https://doi.org/10.3390/app13137534 - 26 Jun 2023
Cited by 1 | Viewed by 1002
Abstract
The activity and disposal options for irradiated Be12Ti were assessed for the HCPB DEMO blankets making use of a code system that enables performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT [...] Read more.
The activity and disposal options for irradiated Be12Ti were assessed for the HCPB DEMO blankets making use of a code system that enables performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. The dedicated full-scale geometry MCNP model of the 11.25 degree HCPB DEMO torus was adapted to the requirements for the coupled 3D neutron transport and activation calculations. Special attention was paid to the use in the activation calculations of the commercial materials containing technological impurities. This has a crucial effect on the results and the impurities must be accounted for in any nuclear safety analyses. The short-term activity is formed by the radionuclides produced through the activation of Be and Ti nuclei and the long-term activity is formed by the products of the neutron irradiation taking place on the impurities. A prerequisite for the disposal or the recycling of the irradiated Be12Ti is its deep detritiation; otherwise, the very high-tritium activity would fully prevent any attempt for its treatment. The most preferable is the use of the Be12Ti with the composition including less material impurities, especially uranium. There could be the option to dispose the Be12Ti intermediate-level wastes in the French repository after 1 year of cooling, assuming the detailed control of the impurities that fulfil the French authority requirements. The USA near-ground repositories could be an alternative to the European sites. The recycling of the irradiated Be12Ti must first be elaborated and approved to ensure its treatment in a safe and efficient way. The remote handling technique must be developed for the re-fabrication of the Be12Ti blocks. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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20 pages, 8199 KiB  
Article
Neutronic Activity for Development of the Promising Alternative Water-Cooled DEMO Concepts
by Pavel Pereslavtsev, Francisco Alberto Hernández, Ivo Moscato and Jin Hun Park
Appl. Sci. 2023, 13(13), 7383; https://doi.org/10.3390/app13137383 - 21 Jun 2023
Cited by 4 | Viewed by 1137
Abstract
An emerging breeding blanket that fulfills performance criteria, meets the safety requirements, and is reliable enough to meet the plant availability is a challenging issue that assumes complex studies involving numerous neutronic analyses based on the Monte Carlo simulations with MCNP code. Two [...] Read more.
An emerging breeding blanket that fulfills performance criteria, meets the safety requirements, and is reliable enough to meet the plant availability is a challenging issue that assumes complex studies involving numerous neutronic analyses based on the Monte Carlo simulations with MCNP code. Two different concepts are now candidates to be implemented as a driver blanket for DEMO fusion reactor: WCLL (Water-Cooled Lithium Lead) and HCPB (Helium-Cooled Pebble Bed). The current R&D work within the EUROfusion DEMO project is concentrated on a search for the new water-cooled blanket layouts: a deep upgrade of the WCLL blanket to ensure a sufficient tritium breeding capability and an elaboration of the hybrid concept coupling technological advantages of water coolant, lead neutron multiplier, and ceramic breeder. To this end, very detailed, fully heterogeneous MCNP geometry models were developed for the newest designs of the WCLL-db (WCLL-double bundle) and WLCB (Water-cooled liquid Lead Ceramic Breeder) DEMO blankets to verify the new engineering solutions. This makes rigorous calculations possible to find an optimal breeder blanket layout. The basic response, tritium breeding ratio (TBR), was assessed for both concepts, and it appeared to be TBR = 1.16 for the WCLL-db and TBR ≤ 1.13 for the WLCB DEMOs, respectively. Several geometry layouts of the WLCB breeder blanket were investigated to reach the TBR sufficient for a sustainable tritium fuel cycle. Two promising novel solutions were suggested to enhance the tritium breeding performance of the WLCB blanket and to achieve TBR ≥ 1.16: heavy water coolant and an advanced breeder ceramic. Various nuclear safety aspects of the technologies utilized in both blanket concepts are addressed in this work to facilitate engineering decisions aimed at the consolidated blanket design for the upcoming DEMO reactor. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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15 pages, 5764 KiB  
Article
Impurity Behavior in Plasma Recovery after a Vacuum Failure in the Experimental Advanced Superconducting Tokamak
by Zihang Zhao, Ling Zhang, Ruijie Zhou, Yang Yang, Wenmin Zhang, Yunxin Cheng, Shigeru Morita, Ang Ti, Ailan Hu, Zhen Sun, Fengling Zhang, Weikuan Zhao, Zhengwei Li, Yiming Cao, Guizhong Zuo and Haiqing Liu
Appl. Sci. 2023, 13(7), 4338; https://doi.org/10.3390/app13074338 - 29 Mar 2023
Cited by 1 | Viewed by 1419
Abstract
After a vacuum failure in a tokamak, plasma runaway or plasma disruptions frequently occur during plasma recovery, causing difficulties in rebuilding a well-confined collisional plasma. In this work, the impurity behavior during plasma recovery after a vacuum failure in the 2019 spring campaign [...] Read more.
After a vacuum failure in a tokamak, plasma runaway or plasma disruptions frequently occur during plasma recovery, causing difficulties in rebuilding a well-confined collisional plasma. In this work, the impurity behavior during plasma recovery after a vacuum failure in the 2019 spring campaign of the Experimental Advanced Superconducting Tokamak (EAST) was studied by analyzing the spectra recorded by fast-time-response extreme ultraviolet (EUV) spectrometers with 5 ms/frame. During the plasma current ramp-up in recovery discharges, a high content of the low-Z impurities of oxygen and carbon was found, i.e., dozens of times higher than that of normal discharges, which may have caused the subsequent runaway discharges. The electron temperature in the recovery discharge may have dropped to less than 75 eV when the collisional plasma quenched to the runaway status, based on the observable impurity ions in the two cases. Therefore, the lifetime of collisional plasma in the recovery discharge, τc, was deduced from the lifetime of H- and He-like oxygen and carbon ions identified from EUV spectra. It was found that, after several discharges with real-time lithium granule injection, the runaway electron flux and O+ influx reduced to 45% and 20%, respectively. Meanwhile, the lifetime of confined plasma was extended from 113 ms to 588 ms, indicating the effective suppression of impurities and runaway electrons and improvement in plasma performance by real-time lithium granule injection. The results in this work provide valuable references for the achievement of first plasma in future superconducting fusion devices such as ITER and CFETR. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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16 pages, 6706 KiB  
Article
Engineering Design of the European DEMO HCPB Breeding Blanket Breeder Zone Mockup
by Guangming Zhou, Joerg Rey, Francisco A. Hernández, Ali Abou-Sena, Martin Lux, Frederik Arbeiter, Georg Schlindwein and Florian Schwab
Appl. Sci. 2023, 13(4), 2081; https://doi.org/10.3390/app13042081 - 6 Feb 2023
Cited by 2 | Viewed by 2365
Abstract
The Helium Cooled Pebble Bed (HCPB) breeding blanket is one of the two driver-blanket candidates for the European fusion demonstration power plant (EU DEMO) within the framework of the EUROfusion Consortium. As the EU DEMO program is going, testing of mockups becomes increasingly [...] Read more.
The Helium Cooled Pebble Bed (HCPB) breeding blanket is one of the two driver-blanket candidates for the European fusion demonstration power plant (EU DEMO) within the framework of the EUROfusion Consortium. As the EU DEMO program is going, testing of mockups becomes increasingly important. In this article, the engineering design of a first-ever breeder zone mockup of the EU DEMO HCPB breeding blanket is reported. The mockup will be tested in the high-pressure, high temperature, helium facility (HELOKA) at Karlsruhe Institute of Technology. This mockup will act as a test rig to validate heat transfer correlations, CFD software, and thermal hydraulics systems codes. As pressure equipment, the mockup shall conform to the latest European Union Pressure Equipment Directive 2014/68/EU. The design description, rationale and test matrix, and corresponding analyses are discussed and presented. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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13 pages, 3953 KiB  
Article
Fast Ion Speed Diffusion Effect on Distributions of Fusion Neutrons
by Pavel Goncharov
Appl. Sci. 2023, 13(3), 1701; https://doi.org/10.3390/app13031701 - 29 Jan 2023
Viewed by 1321
Abstract
Velocity distributions of fuel nuclei enter the formulae for distributions of products of fusion reactions in plasma. The formulae contain multiple integration, which is a computationally heavy task. Therefore, simplifications of the integrand are advantageous. One of possible simplifications is the use of [...] Read more.
Velocity distributions of fuel nuclei enter the formulae for distributions of products of fusion reactions in plasma. The formulae contain multiple integration, which is a computationally heavy task. Therefore, simplifications of the integrand are advantageous. One of possible simplifications is the use of closed-form analytical distributions of fast deuterons and tritons, accounting for slowing down and pitch-angle scattering and neglecting the speed diffusion. The plausibility of such a model has been studied from the viewpoint of its influence on the calculated spectra of fusion neutrons. Calculations have shown that the speed diffusion effect on suprathermal ion distribution tails does not significantly alter the qualitative behaviour of energy and angle distributions of fusion products in a beam-heated plasma. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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13 pages, 7082 KiB  
Article
Development of a Set of Synthetic Diagnostics for the Confrontation between 2D Transport Simulations and WEST Tokamak Experimental Data
by Ivan Kudashev, Anna Medvedeva, Manuel Scotto d’Abusco, Nicolas Fedorszak, Stefano Di Genova, Vladislav Neverov and Eric Serre
Appl. Sci. 2022, 12(19), 9807; https://doi.org/10.3390/app12199807 - 29 Sep 2022
Cited by 5 | Viewed by 1982
Abstract
Transport codes are frequently used for describing fusion plasmas with the aim to prepare tokamak operations. Considering novel codes, such as SolEdge3X-HDG, synthetic diagnostics are a common technique used to validate new models and confront them with experimental data. The purpose of this [...] Read more.
Transport codes are frequently used for describing fusion plasmas with the aim to prepare tokamak operations. Considering novel codes, such as SolEdge3X-HDG, synthetic diagnostics are a common technique used to validate new models and confront them with experimental data. The purpose of this study is to develop a set of synthetic diagnostics, starting from bolometer and visible cameras for the WEST tokamak, in order to compare the code results with the experimental data. This research is done in the framework of Raysect and Cherab Python libraries. This allows us to process various synthetic diagnostics in the same fashion in terms of 3D ray tracing with volume emitters developed specifically for fusion plasmas. We were able to implement the WEST tokamak model and the design of bolometer and visible cameras. Synthetic signals, based on full-discharge WEST plasma simulation, were used for to compare the SolEdge3X-HDG output plasma with experimental data. The study also considers the optical properties of the plasma-facing components (PFCs) and their influence on the performance of diagnostics. The paper shows a unified approach to synthetic diagnostic design, which will be further extended to cover the remaining diagnostics on the WEST tokamak. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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27 pages, 10823 KiB  
Article
Neutral Beam Coupling with Plasma in a Compact Fusion Neutron Source
by Eugenia Dlougach, Alexander Panasenkov, Boris Kuteev and Arkady Serikov
Appl. Sci. 2022, 12(17), 8404; https://doi.org/10.3390/app12178404 - 23 Aug 2022
Cited by 6 | Viewed by 2082
Abstract
FNS-ST is a fusion neutron source project based on a spherical tokamak (R/a = 0.5 m/0.3 m) with a steady-state neutron generation of ~1018 n/s. Neutral beam injection (NBI) is supposed to maintain steady-state operation, non-inductive current drive and neutron production in [...] Read more.
FNS-ST is a fusion neutron source project based on a spherical tokamak (R/a = 0.5 m/0.3 m) with a steady-state neutron generation of ~1018 n/s. Neutral beam injection (NBI) is supposed to maintain steady-state operation, non-inductive current drive and neutron production in FNS-ST plasma. In a low aspect ratio device, the toroidal magnetic field shape is not optimal for fast ions confinement in plasma, and the toroidal effects are more pronounced compared to the conventional tokamak design (with R/a > 2.5). The neutral beam production and the tokamak plasma response to NBI were efficiently modeled by a specialized beam-plasma software package BTR-BTOR, which allowed fast optimization of the neutral beam transport and evolution within the injector unit, as well as the parametric study of NBI induced effects in plasma. The “Lite neutral beam model” (LNB) implements a statistical beam description in 6-dimensional phase space (106–1010 particles), and the beam particle conversions are organized as a data flow pipeline. This parametric study of FNS-ST tokamak is focused on the beam-plasma coupling issue. The main result of the study is a method to achieve steady-state current drive and fusion controllability in beam-driven toroidal plasmas. LNB methods can be also applied to NBI design for conventional tokamaks. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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19 pages, 1868 KiB  
Article
Numerical Modeling of Individual Plasma Dynamic Characteristics of a Light-Erosion MPC Discharge in Gases
by Victor V. Kuzenov, Sergei V. Ryzhkov and Aleksey Yu. Varaksin
Appl. Sci. 2022, 12(7), 3610; https://doi.org/10.3390/app12073610 - 1 Apr 2022
Cited by 23 | Viewed by 2000
Abstract
A mathematical model is formulated, and a numerical study of magneto-plasma compressor (MPC) discharges in gases for a wide range of changes in the main electrical parameters and the characteristics of the surrounding gas environment is carried out. The performed calculations showed, depending [...] Read more.
A mathematical model is formulated, and a numerical study of magneto-plasma compressor (MPC) discharges in gases for a wide range of changes in the main electrical parameters and the characteristics of the surrounding gas environment is carried out. The performed calculations showed, depending on the role of one or another heating mode (Ohmic, transient, and plasma dynamic), three different types of quasi-stationary spatial distributions of plasma parameters, which can be used to judge the features of the emerging structures and the dynamics of plasma propagation, and, therefore, to speak about the modes of discharge. The features of the radiation plasma dynamic structures and the change in the main parameters of the plasma of an MPC discharge in the transient regime are considered. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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26 pages, 2165 KiB  
Article
A Model to Simulate Gas Dissolution into/through Metals and Its Application to Deuterium in a 316L Steel Chamber with Pb-Li in a Quasi-2D Geometry
by Tiago Pomella Lobo, Ester Diaz-Alvarez and Laëtitia Frances
Appl. Sci. 2022, 12(5), 2523; https://doi.org/10.3390/app12052523 - 28 Feb 2022
Cited by 1 | Viewed by 2098
Abstract
Liquid lead-lithium in eutectic proportions (Pb-Li) is a candidate material for Breeding Blankets (BBs) in future Fusion Power Plants (FPP). BB design depends on the diffusivity and Sieverts’ constant (solubility) of tritium in this alloy, but literature reports a large scattering of measurements [...] Read more.
Liquid lead-lithium in eutectic proportions (Pb-Li) is a candidate material for Breeding Blankets (BBs) in future Fusion Power Plants (FPP). BB design depends on the diffusivity and Sieverts’ constant (solubility) of tritium in this alloy, but literature reports a large scattering of measurements for these values. A model was developed to address one possible source of this scattering in static experiments, i.e., non-negligible loss of hydrogen gas through steel walls of containers. This model simulates the dissolution of gases into, and their diffusion through, metallic barriers for diffusivity and Sieverts’ constant as inputs. When implemented, it can be used to compute the pressure decrease in a metallic chamber, and comparison of simulated curves with experimental ones allows for estimates of the diffusivity and Sieverts’ constant. This approach was used to estimate these coefficients for deuterium in stainless steel, using experiments performed with a 316L steel chamber from an existing facility (the Vacuum Sieve Tray setup) and simulations in a quasi-2D representation of this chamber. This validated the model, which was then used to simulate the chamber containing Pb-Li, as a means of planning for future experiments. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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12 pages, 935 KiB  
Article
Study of RF Stray Currents in ITER Neutral Beam Test Facilities
by Riccardo Casagrande, Alberto Maistrello, Marco De Nardi, Mattia Dan and Mauro Recchia
Appl. Sci. 2021, 11(23), 11126; https://doi.org/10.3390/app112311126 - 24 Nov 2021
Cited by 5 | Viewed by 2126
Abstract
The operation of SPIDER (Source for the Production of Ions of Deuterium Extracted from Radio-frequency plasma), full-scale prototype of ITER NBI (Neutral Beam Injector) radio-frequency ion source, pointed out deleterious effects caused by stray Radio-Frequency (RF) currents flowing in the electrical equipment not [...] Read more.
The operation of SPIDER (Source for the Production of Ions of Deuterium Extracted from Radio-frequency plasma), full-scale prototype of ITER NBI (Neutral Beam Injector) radio-frequency ion source, pointed out deleterious effects caused by stray Radio-Frequency (RF) currents flowing in the electrical equipment not included in the RF power system. MITICA (Megavolt ITER Injector and Concept Advancement), the full-scale prototype of ITER NBI, is characterized by a similar design in terms of layout of the power supplies and connections to the beam source; thus, it is expected to be subject to the RF stray currents problem. SPIDER RF system is composed of four RF generators, four coaxial lines and four RF loads. Each RF generator is rated for operation at 200 kW in the frequency range 0.9 ÷ 1.1 MHz. The power is delivered to the four loads via as many RF coaxial lines, housed inside a multiconductor transmission line. Each load consists of a capacitive matching network and two plasma drivers in series. Due to the presence of stray connections at the generator and beam-source side (e.g., parasitic capacitances), unwanted RF currents can flow through alternative paths and negatively affect the components not intended for transmission of RF power, the output stages of power supplies and several diagnostics installed in the High-Voltage Deck (HVD) and at the beam source. This paper presents the development of a circuital model used to estimate the RF stray currents in SPIDER electrical system; the understanding of this phenomenon and the development of a model with predictive capabilities is fundamental for the assessment of the performance of both SPIDER and MITICA and, in general, of alternative RF system layouts with respect to the stray currents issue. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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11 pages, 1955 KiB  
Article
Selection of Fuel Isotope Composition in Heating Injectors of the FNS-ST Compact Fusion Neutron Source
by Sergey Ananyev, Alexey Dnestrovskij and Andrei Kukushkin
Appl. Sci. 2021, 11(16), 7565; https://doi.org/10.3390/app11167565 - 18 Aug 2021
Cited by 4 | Viewed by 1818 | Correction
Abstract
For the FNS-ST compact neutron source, the dependence of the neutron yield on the tritium content in the bulk plasma is analyzed for the operation of the heating injectors with different isotope compositions of the neutral beams. Self-consistent simulations of the FNS-ST operating [...] Read more.
For the FNS-ST compact neutron source, the dependence of the neutron yield on the tritium content in the bulk plasma is analyzed for the operation of the heating injectors with different isotope compositions of the neutral beams. Self-consistent simulations of the FNS-ST operating regimes are performed using the SOLPS4.3 and ASTRA codes for different densities of the bulk plasma and diffusion coefficients. The FC-FNS code is used to calculate the required fluxes of the fuel components into the plasma provided by different injection systems: the pellet injectors and the neutral beams. In simulations, the plasma density is varied in the range ne = (7–10) × 1019 m−3, and the ratio of the diffusivity to the heat conductivity in the range D/χe = 0.2–0.6. For the scenarios with the D + T or D beams, in the window of the operating parameters, the maximum possible fractions of tritium in the bulk plasma are calculated, and the corresponding neutron yields are obtained. For the regimes with the maximum neutron yield (4.5–5.5) × 1017 s−1, the accumulation of tritium at the site (up to 550 g) is calculated for different heating beams. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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16 pages, 7173 KiB  
Article
Statistical Analysis of Tritium Breeding Ratio Deviations in the DEMO Due to Nuclear Data Uncertainties
by Jin Hun Park, Pavel Pereslavtsev, Alexandre Konobeev and Christian Wegmann
Appl. Sci. 2021, 11(11), 5234; https://doi.org/10.3390/app11115234 - 4 Jun 2021
Cited by 6 | Viewed by 2457
Abstract
For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. [...] Read more.
For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. The uncertainty of the TBR in the neutronic simulations includes many different aspects such as the uncertainty due to the simplification of the geometry models used, the uncertainty of the reactor layout and the uncertainty introduced due to neutronic calculations. The last one can be reduced by applying high fidelity Monte Carlo simulations for TBR estimations. Nevertheless, these calculations have inherent statistical errors controlled by the number of neutron histories, straightforward for a quantity such as that of TBR underlying errors due to nuclear data uncertainties. In fact, every evaluated nuclear data file involved in the MCNP calculations can be replaced with the set of the random data files representing the particular deviation of the nuclear model parameters, each of them being correct and valid for applications. To account for the uncertainty of the nuclear model parameters introduced in the evaluated data file, a total Monte Carlo (TMC) method can be used to analyze the uncertainty of TBR owing to the nuclear data used for calculations. To this end, two 3D fully heterogeneous geometry models of the helium cooled pebble bed (HCPB) and water cooled lithium lead (WCLL) European DEMOs were utilized for the calculations of the TBR. The TMC calculations were performed, making use of the TENDL-2017 nuclear data library random files with high enough statistics providing a well-resolved Gaussian distribution of the TBR value. The assessment was done for the estimation of the TBR uncertainty due to the nuclear data for entire material compositions and for separate materials: structural, breeder and neutron multipliers. The overall TBR uncertainty for the nuclear data was estimated to be 3~4% for the HCPB and WCLL DEMOs, respectively. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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19 pages, 7714 KiB  
Article
Development of a Component-Level Hydrogen Transport Model with OpenFOAM and Application to Tritium Transport Inside a DEMO HCPB Breeder
by Volker Pasler, Frederik Arbeiter, Christine Klein, Dmitry Klimenko, Georg Schlindwein and Axel von der Weth
Appl. Sci. 2021, 11(8), 3481; https://doi.org/10.3390/app11083481 - 13 Apr 2021
Cited by 3 | Viewed by 2139
Abstract
This work continues the development of a numerical model to simulate transient tritium transport on the breeder zone (BZ) level for the EU helium-cooled pebble bed (HCPB) concept for DEMO. The basis of the model is the open-source field operation and manipulation framework, [...] Read more.
This work continues the development of a numerical model to simulate transient tritium transport on the breeder zone (BZ) level for the EU helium-cooled pebble bed (HCPB) concept for DEMO. The basis of the model is the open-source field operation and manipulation framework, OpenFOAM. The key output quantities of the model are the tritium concentration in the purge gas and in the coolant and the tritium inventory inside the BZ structure. New model features are briefly summarized. As a first relevant application a simulation of tritium transport for a single pin out of the KIT HCPB design for DEMO is presented. A variety of scenarios investigates the impact of the permeation regime (diffusion-limited vs. surface-limited), of an additional hydrogen content of 300 Pa H2 in the purge gas, of the released species (HT vs. T2), and of the choice of species-specific rate constants (recombination constant of HT set twice as for H2 and T2). The results indicate that the released species plays a minor role for permeation. Both permeation and inventory show a considerable dependence on a possible hydrogen addition in the purge gas. An enhanced HT recombination constant reduces steel T inventories and, in the diffusion-limited case, also permeation significantly. Scenarios with 80 bar vs. 2 bar purge gas pressure indicate that purge gas volumetric flow is decisive for permeation. Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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Review

Jump to: Research

13 pages, 291 KiB  
Review
Magneto-Inertial Fusion and Powerful Plasma Installations (A Review)
by Sergei V. Ryzhkov
Appl. Sci. 2023, 13(11), 6658; https://doi.org/10.3390/app13116658 - 30 May 2023
Cited by 11 | Viewed by 2466
Abstract
A review of theoretical and experimental studies in the field of compression and heating of a plasma target in an external magnetic field, which has recently been called magneto-inertial fusion (MIF), has been carried out. MIF is a concept of magnetically driven inertial [...] Read more.
A review of theoretical and experimental studies in the field of compression and heating of a plasma target in an external magnetic field, which has recently been called magneto-inertial fusion (MIF), has been carried out. MIF is a concept of magnetically driven inertial fusion that involves the magnetization of fuel, laser pre-heating, and magnetic implosion to create fusion conditions. An analysis of the current state of work on the implosion of magnetized targets and the effect of an external magnetic field on the main plasma parameters and system characteristics is presented. Questions regarding the numerical simulation of experiments on the magnetic-inertial confinement of plasma are touched upon. Particular attention is paid to two promising areas of MIF—with plasma jets and with a laser driver (laser beams). Full article
(This article belongs to the Special Issue Advances in Fusion Engineering and Design)
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