Development Of Fusion Reactors Research Articles

Conceptual design of poloidal horseshoe limiter layout for JA DEMO

In the development of the JA DEMO fusion reactor, the proposed poloidal horseshoe limiter aims to mitigate the surface heat load localized on the breeding blanket. The limiter protrudes from the first wall toward the plasma to shade the blanket from the charged particles, which account for the majority of the surface heat load. To concretize the limiter design, this study presents a limiter layout that implements a concept that allows for sufficient cooling and remote maintenance. The limiter shape was determined by evaluating the surface heat load using 3-D magnetic field line tracing. A fully poloidal limiter with a width of 300 mm, which corresponds to 2.4 % of the FW area, was selected. The limiter heating, consisting of surface heat load and nuclear heating, was evaluated to discuss the required cooling, and the concept of a limiter cassette and modules for cooling was proposed. For the limiter replacement method, the outboard limiter is designed to be extracted from the upper port with the mounting segment. The inboard limiter is then grasped and transported to reach the upper port after the outboard limiter extraction. Both the inboard and outboard limiter are expected to be replaced without affecting other in-vessel components in this study.

Physics design of 14 MeV neutron generator facility at the Institute for Plasma Research

A high energy and high yield neutron source is a prime requirement for technological studies related to fusion reactor development. It provides a high-energy neutron environment for small-scale fusion reactor components research and testing such as tritium breeding, shielding, plasma-facing materials, reaction cross-section data study for fusion materials, etc. Along with ITER participation, the Institute of Plasma Research, India is developing an accelerator-based 14 MeV neutron source with a yield of 1012 n s−1. The design of the source is based on the deuterium–tritium fusion reaction. The deuterium beam is accelerated and delivered to the tritium target to generate 14 MeV neutrons. The deuterium beam energy and tritium availability in the tritium target are the base parameters of the accelerator-based neutron source design. The paper gives the physics design of the neutron generator facility of the Institute for Plasma Research. It covers the requirements, design basis, and physics parameters of the neutron generator. As per the analytical results generator can produce more than 1 × 1012 n s−1 with a 110 keV D+ ion beam of 10 mA and a minimum 5 Ci tritium target. However, the detailed simulation with the more realistic conditions of deuteron ion interaction with the tritium titanium target shows that the desired results cannot be achieved with 110 keV. The safe limit of the ion energy should be 300 keV as per the simulation. At 300 keV ion energy and 20 mA current, it reaches 1.6 × 1012 n s−1. Moreover, it was found that to ensure sufficiently long operation time a tritium target of more than 20 Ci should be used. The scope of the neutron source is not limited to the fusion reactor research studies, it is extended to other areas such as medical radioisotopes research, semiconductor devices irradiations, and many more.

Effect of tungsten doping on the irradiation resistance of FeCrV-based refractory medium entropy alloy for potential nuclear applications

Given the increasing demand for energy and the rising cost of fossil fuels, the development of nuclear fission and fusion reactors has become a key priority. Recently, High Entropy Alloys (HEAs) and Medium Entropy Alloys (MEAs) have emerged as promising materials for use in nuclear reactors due to their unique properties, which include significant lattice distortion, high configurational entropy, reduced diffusivity, and the cocktail effect. In this study, we designed and analysed a series of FeCrV-based Refractory Medium-entropy Alloys (RMEAs) as structural materials in fusion reactors, using Density Functional Theory. By calculating the empirical parameters, we have verified the solid solution's structure stability of the FeCrV-based RMEA. We examined the mechanical, electronic, and irradiation resistance properties of these alloys and found that the incorporation of tungsten dopants could enhance both the mechanical properties and irradiation resistance of the materials, with the optimal doping level being approximately 8% of Tungsten. Furthermore, electronic structure calculations were performed to elucidate the underlying mechanisms. This knowledge can guide the development of new HEA/MEA materials for use in nuclear reactors.

Development of steady-state fusion reactor by Helical Fusion

The Helical Fusion group (Helical Fusion Co., Ltd. in Japan and Helical Fusion USA, Inc.) is developing a 50 MWe-class steady-state helical fusion reactor, which is a kind of stellarator called heliotron composed of two continuous helical coils similar to the large helical device and is operated without plasma current. HESTIA plays the role of the fusion pilot plant. The first-of-a-kind fusion power plant that would follow HESTIA will be a 100 MWe-class steady-state helical fusion reactor. After more than one year of continuous operation, maintenance will be completed within 3 months, aiming for an availability larger than 80%. High-temperature superconducting magnet coils are adopted to generate 8 T of the magnetic field at the helical coil center. HESTIA is a deuterium-tritium fusion reactor, where tritium is self-produced using liquid metal blanket systems. The first wall of the liquid metal blanket including the divertor strike zone is covered and protected by liquid metal free-surface flow, and therefore, individual divertor systems are not required in HESTIA. Electron cyclotron heating is adopted for plasma heating. Since plasma current drive is not required, HESTIA can be operated at a low fusion gain of ∼13, and steady-state operation is principally possible on the order of a year. After a few years of individual development phases, a prototype device is planned to be constructed and operated for the integrated demonstration before starting the construction of HESTIA.

High Heat Flux Testing of Graded W-Steel Joining Concepts for the First Wall

The realization of the first wall (FW), which is composed of a protective tungsten (W) armor covering the structural steel material, is a critical challenge in the development of future fusion reactors. Due to the different coefficients of thermal expansion (CTE) of W and steel, the direct joining of them results in cyclic thermal stress at their bonding seam during the operation of the fusion reactor. To address this issue, this study benchmarks two joining concepts. The first concept uses an atmospheric plasma sprayed graded interlayer composed of W/steel composites with a varying content of W and steel to gradually change the CTE. The second concept uses a spark plasma sintered graded interlayer. Furthermore, in order to benchmark these concepts, a directly bonded W-steel reference joint as well as a W-steel joint featuring a vanadium interlayer were also tested. These joints were tested under steady-state high heat flux cyclic loading, starting from a heat flux of 1 MW/m2 up to 4.5 MW/m2, with stepwise increments of 0.5 MW/m2. At each heat flux level, 200 thermal cycles were performed. The joints featuring a sintered graded interlayer survived only until 1.5 MW/m2 of loading, while the joint featuring plasma sprayed graded interlayer and V interlayer survived until 3 MW/m2.

Filamentary transport in global edge-SOL simulations of ASDEX Upgrade

The understanding and the predictive capability for turbulence in the plasma edge and scrape-off layer (SOL) are crucial for the development of magnetic confinement fusion reactors. To this end, we characterise turbulent transport across the edge and SOL of the diverted ASDEX Upgrade tokamak in attached L-mode conditions by means of validated, global simulations. The collisionality is controlled by the divertor neutrals density, as their ionization increases the plasma density and decreases the temperature. The radial E×B particle and heat transport, quantified by effective diffusivities, rises strongly with collisionality. The modest increase in fluctuation amplitudes is not a sufficient explanation. The reason is shown to be the destabilisation of resistive drift-ballooning modes, resulting in larger phase shifts between the pressure and electrostatic potential. The transport varies both radially and poloidally. Due to its ballooning nature, radial transport is close to zero on the inboard mid-plane. On the outboard mid-plane, significant transport is driven in the SOL by large filaments (blobs) with amplitudes of up to 250% of the mean, propagating ballistically from the separatrix to the wall. This non-local transport leads to large radial variations of diffusivities, as they do not necessarily correlate with the local gradient. Ion temperature fluctuations in the plasma edge are shown to be involved in blob seeding at the separatrix, and are the largest in the SOL. Radial diffusivities peak at the top and bottom of the device, since gradients are flatter there due to the flux expansion while the cross-field flow is sustained by streamers—a feature which should be considered in mean-field transport modeling. The increase of SOL E×B transport with collisionality is likely fostered by the simultaneously decreasing radial electric field, resulting from a flattened electron temperature profile. Large amplitude blobs are a hazard for plasma facing components of fusion reactors, but they could be restrained by control of SOL collisionality.

Preliminary investigation of neutron shielding compounds for ARC-class tokamaks

Radiation shielding design is a crucial step in the development of a nuclear fusion reactor. Superconducting magnets are indeed a highly engineered component whose properties are particularly sensitive to irradiation damage. In addition, novel magnet technologies, such as High Temperature Superconductors (HTS), and the attractiveness of compact tokamak designs (like Affordable Robust Compact - ARC - reactor), require an even higher precision in the shielding design. In this work most promising radiation shielding solutions for compact tokamaks are explored. The study applies a Monte Carlo particles transport technique assisted by the OpenMC code. A D-shaped model is built and ARC reactor core is considered as main case study. A detailed analysis of the radiation environment is carried out. The study then assesses the most promising compounds on the basis of most suitable cross sections, material shielding properties and space constraints. In addition, the work addresses the problem of energy deposition on the shield providing an analytical model for active cooling necessities. Results highlighted the necessity of heavy elements and high density compounds as a considerable photon flux is generated in neutron-material interactions. Tungsten borides (WxBy) ceramics seem to be particularly promising for shielding radiation in a compact configuration.

Development of Poloidal Horseshoe Limiter Concept for JA DEMO

In the development of a DEMO fusion reactor, mitigating the heat load deposited on the breeding blanket at steady-state operation is one of the challenging issues because of the localized high heat flux brought by charged particles and low thermal conductivity of the reduced activation ferritic/martensite (RAFM) steel structural material. This study proposes a concept of limiter for JA DEMO; a horseshoe-shaped limiter which is continuous poloidally in the first wall (FW) except for the diverter area, and discretized toroidally. The limiter is protruded from the FW to shade the blankets from the charged particles, which move along the magnetic field lines. To enhance the heat removal capability and neutron irradiation tolerance, a tungsten mono-block using RAFM heat sink was applied as the plasma-facing unit (PFU) on the limiters. The heat removal capability of the limiter’s PFU was evaluated as 4.6 MW/m2. The limiter shape satisfying this value was determined basing on the utilization of heat load evaluation with 3-D magnetic field line tracing. Three limiter shape parameters; protrusion height, protrusion width and the FW set-back, were investigated to understand the variation of charged particle load distribution. Both cylindrical and box-shaped blanket concepts were considered, and design windows for the limiter shape were acquired to discuss the optimum parameter. The optimum limiter shape fulfilling the requirement of steady-state operation has been defined, and the limiter’s occupied area amounts to 1.2% of the FW area for both blanket options.

A mini review on a study of Helium behavior for fusion reactor development

A mini review on a study of Helium behavior for fusion reactor development

Simulations of charge collection of a gallium nitride based pin thin-film neutron detector

The development of new fast neutron reactors and nuclear fusion reactors requires new neutron detectors in extreme environments. Due to its wide bandgap (3.4 eV) and radiation resistance capability, gallium nitride (GaN) is a candidate for neutron detection in extreme environments. This study introduces a novel simulation method of charge collection efficiency (CCE) for GaN pin thin-film neutron detector based on the Hecht equation and Monte Carlo simulation. A modified 2-carrier Hecht equation is used to simulate the CCE of the detector with a different depth depletion region. After obtaining the neutron energy deposition distribution in the sensitive volume of the detector, the Hecht equation is used to calculate the charge collection efficiency at different positions of the detector under a uniform electric field. The maximum relative error between the simulated CCE and the experimental CCE value is about 6.3%.

Low cost, simpler HTS cable conductors for fusion energy systems

The cost and complexity of large, high-field superconducting magnet modules and related subsystems comprise 30% to 60% of the fusion reactor core capital cost. The strategic plan for the U.S. burning plasma research, the Fusion Energy Sciences Committee Report (FESAC) “Power the Future: Fusion and Plasmas’', and 2021 NASEM report “Key Goals and Innovations needed for a U.S. Fusion Pilot Plant” recommends that the U.S. pursue innovative science and technology to enable construction of a Fusion Pilot Plant (FPP) that produces net electricity from fusion at reduced capital cost. To achieve this, a novel combination of lower-cost high temperature superconductors (HTS) in cable configurations with co-wound reinforcement for higher current density are being investigated using a simplified construction strategy to produce compact stable coils. They would be capable of generating 20 T at up to 10-20 K. Small-scale, inexpensive test coils and prototypes will help develop each feature and validate cabled conductor design models. The near term goal is to validate engineering approaches, scientific models and fabrication capabilities applicable to fusion reactor development such as U.S. fusion nuclear science facility (FNSF), sustained high-power density tokamak facility (SHPD) and FPP designs. The design options include lower-cost, high-strength, quench resistant REBCO or Bi-2212 cables in an all metal coil design that simplifies HTS coil construction and quench protection system, with co-wound reinforcements that integrate stress management in HTS cable design and provides thermal mass to help prevent quench damage.

Development of a concept and basis for the DEMO diagnostic and control system

An initial concept for the plasma diagnostic and control (D&C) system has been developed as part of European studies towards the development of a demonstration tokamak fusion reactor (DEMO). The main objective is to develop a feasible, integrated concept design of the DEMO D&C system that can provide reliable plasma control and high performance (electricity output) over extended periods of operation. While the fusion power is maximized when operating near to the operational limits of the tokamak, the reliability of operation typically improves when choosing parameters significantly distant from these limits. In addition to these conflicting requirements, the D&C development has to cope with strong adverse effects acting on all in vessel components on DEMO (harsh neutron environment, particle fluxes, temperatures, electromagnetic forces, etc.). Moreover, space allocation and plasma access are constrained by the needs for first wall integrity and optimization of tritium breeding. Taking into account these boundary conditions, the main DEMO plasma control issues have been formulated, and a list of diagnostic systems and channels needed for plasma control has been developed, which were selected for their robustness and the required coverage of control issues. For a validation and refinement of this concept, simulation tools are being refined and applied for equilibrium, kinetic and mode control studies.

Theoretical simulation of the closed currents near non-uniformly strongly heated surface of tungsten due to thermo-emf

A problem of surface melting under the impact of plasmas is one of the most important in the development of future magnetic confinement fusion reactors. The expected high heat loads can lead to melting of tungsten chosen as the material of a divertor and the first wall of ITER. The tungsten melt can move under the action of forces and quickly deform a surface. This article shows that at high temperatures, a tungsten vapor can be considered as a conductive material with conductivity high enough to be taken into account during simulation of current through a sample. We describe the mechanism of current generation by the thermo-emf due to non-uniform heating of the vapor/condensed substance interface without external sources of charges. Even without direct contact with the external plasma, the density of this current is high enough to cause noticeable movement of the melt in an external magnetic field. Simulation shows that the melt moving observed at BETA may be caused not by a current of an electron beam but by non-uniformity of heating and can be reproduced by any other similar heat source in the external magnetic field.

Recent results from deuterium experiments on the large helical device and their contribution to fusion reactor development

In recent deuterium experiments on the large helical device (LHD), we succeeded in expanding the temperature domain to higher regions for both electron and ion temperatures. Suppression of the energetic particle driven resistive interchange mode (EIC) by a moderate electron temperature increase is a key technique to extend the high temperature domain of LHD plasmas. We found a clear isotope effect in the formation of the internal transport barrier in high temperature plasmas. A new technique to measure the hydrogen isotope fraction was developed in the LHD in order to investigate the behavior of the isotope mixing state. The technique revealed that the non-mixing and the mixing states of hydrogen isotopes can be realized in plasmas. In deuterium plasmas, we also succeeded in simultaneously realizing the formation of the edge transport barrier (ETB) and the divertor detachment. It is found that resonant magnetic perturbation plays an important role in the simultaneous formation of the ETB and the detachment. Contributions to fusion reactor development from the engineering point of view, i.e. negative-ion based neutral beam injector research and the mass balance study of tritium, are also discussed.

Preliminary MHD pressure drop analysis for the prototypical WCLL TBM with RELAP5/MOD3.3

In fusion reactor development, the design of the breeding blanket (BB) represents an important technical challenge due to the extreme operative conditions that the component must withstand. One of the critical issues of a liquid metal BB is the occurring of magnetohydrodynamic (MHD) effects within the piping network that transports the electro-conductive fluid. Predicting MHD phenomena is fundamental for a reliable and efficient blanket design and computational tools are indispensable to achieve a fusion-relevant physical analysis. For this reason, a dedicated MHD module for the System Thermal-Hydraulic (STH) code RELAP5/MOD3.3 (RELAP5) is under development at Sapienza University of Rome. In its current state, RELAP5 can evaluate MHD pressure losses in many common layouts for BB piping system. In a previous work of the Authors, the verification and validation (V&V) procedure of the MHD module has been discussed. In this paper, the code is used to assess the MHD pressure loss in the Water-Cooled Lithium Lead Test Blanket Module (TBM). Furthermore, the geometry of the TBM box is optimized to reduce the MHD pressure loss and to equalize the flow distribution among the Breeding Units.

Irradiation effects in tungsten—From surface effects to bulk mechanical properties

Advanced materials such as tungsten fibre-reinforced composites allow to overcome severe weaknesses of the baseline materials for plasma-facing components — copper and tungsten. The effect of the fusion environment on the mechanical properties of these materials, e.g. the embrittlement by neutron irradiation, plays a key role for the development of future fusion reactors. To simulate this effect, high-energy ions are used as a substitute for the displacement damage by neutrons. We propose the use of very fine tungsten wire as a possibility of studying the influence of irradiation damage on the mechanical properties. This is possible as they allow full-depth irradiation of almost the entire volume despite the limited penetration depth of ions. Geometrical size effects are mitigated due to the nanoscale microstructure of the wire. In addition, similar wire is used in tungsten fibre-reinforced composites. Thus, the investigation of irradiated wire can directly be used for the prediction of the bulk composite properties. For the proof of this concept tungsten wire with a diameter of 16 μm was electrochemically thinned to 5 μm and irradiated with 20.5 MeV W6+ ions. The mechanical properties were subsequently determined by macroscopic tensile testing. Irradiation to 0.3, 1 and 9 dpa did not lead to a change of the mechanical behaviour. Both strength and ductility, the latter indicated by the reduction of area, were similar to the as-fabricated state.

First evidence of dominant influence of E × B drifts on plasma cooling in an advanced slot divertor for tokamak power exhaust

Addressing power exhaust in tokamaks is presently recognized as one of the major remaining open issues for the development of fusion reactors. At the forefront of this endeavor is the effort to develop an advanced divertor by maximizing dissipation of plasma power and momentum inside the divertor. Here, we demonstrate, for the first time, that the electromagnetic ( E × B ) drifts exert a key influence on plasma and gas dynamics in a new advanced slot divertor in the DIII-D tokamak, named the small angle slot (SAS). SAS leverages the effect of drifts to achieve a highly dissipative divertor with electron temperature T e ≲ 10 eV over a wide range of plasma densities, for ion B × ∇ B away from the divertor, as used for advanced tokamak operation on DIII-D. Modeling with the SOLPS-ITER code shows that for this drift direction, the E × B flow carries particles toward the outer common flux region (CFR) via the private flux region (PFR), reinforcing neutral recycling and enhancing divertor dissipation. In contrast, for the opposite field direction, the E × B flow carries particles away from the outer CFR into the PFR, offsetting the anticipated SAS geometric effects. This finding is an important step in the understanding of the behavior of advanced divertors for achieving a power exhaust solution for fusion reactors.

Pre-conceptual design of EU-DEMO divertor primary heat transfer systems

In the frame of the activities promoted and encouraged by the EUROfusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently addressed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, the Divertor Primary Heat Transfer Systems (PHTSs) are intended to feed coolant to the two main components of the Divertor assembly, namely the Plasma Facing Components (PFCs) and the Cassette Body (CB). Since the DEMO Divertor must withstand high heat flux loads together with a considerable neutron deposited power, very tight tolerances may be allowed to the coolant inlet conditions. Therefore, the design of reliable PHTSs is of the utmost importance towards the development of an EU-DEMO fusion reactor. Within this framework, a study has been jointly carried out by University of Palermo, Ansaldo Nucleare and CREATE to design the Ex-Vessel PHTSs of both the PFCs and the CB for a DEMO reactor equipped with a Helium Cooled Pebble Bed Breeding Blanket. The paper describes criteria and rationale followed with the aim to achieve simple PHTS designs based on the adoption of easy-to-manufacture main components avoiding too much extrapolation from the state-of-art technology. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, pressurizers, piping and pumps are presented and critically discussed, with particular attention to those integration, safety and feasibility constraints that may deeply affect the design of such components. Finally, the evaluation of PHTS key parameters as total pumping power and coolant inventory is reported.

A full and heterogeneous model of the ITER tokamak for comprehensive nuclear analyses

ITER is the flagship fusion project, conceived as an experiment to select and develop the technologies for the first demonstration reactor, DEMO. Nuclear analysis is a core discipline in support of the design, commissioning and operation of the machine. To date, it has been conducted with increasingly detailed partial models, which represented toroidal segments of the tokamak. However, the limitations of this methodology became evident as estimates of quantities relevant to design, safety and operation showed unquantifiable uncertainties, which is a risk. Here, we present a detailed and realistic 360° MCNP model of the ITER tokamak called E-lite. We demonstrate the model’s usability and practicality. Two examples are used to illustrate qualitatively and quantitatively how it solves previously intractable problems with marked benefits for the future nuclear analysis of ITER, with applications to DEMO and future reactors. E-lite constitutes a milestone in the field of nuclear analysis in terms of realism in the evaluation of key quantities. The development of nuclear fusion reactors requires detailed nuclear analyses to ensure they run safely and effectively. These are currently typically done using partial models. Juarez et al. now present a complete 360° model of the ITER tokamak, which allows for more realistic and complete modelling of the system.

Broad Beam-Induced Fragmentation and Joining of Tungsten Oxide Nanorods: Implications for Nanodevice Fabrication and the Development of Fusion Reactors

In this work, for the first time, fragmentation and joining of tungsten oxide (WO3) nanorods induced by a broad ion beam are reported. Although at low energy (5 keV) and moderate ion fluence, nanorods fragment into smaller pieces along the length, at higher ion energies (50–100 keV), a contrary process occurs, which leads to the joining of the nanorods. A state-of-the-art ion–solid interaction simulation, namely, TRI3DYN, has been invoked to explore the possible mechanisms that reveal subtle contributions of surface defects, ion-beam mixing, and sputtering. High-resolution electron microscopy, photoluminescence study, and X-ray photoelectron spectroscopy support the observed results and proposed mechanisms. Such modifications have interesting effects on the electrical conductivity of the nanorod assembly. The change in sample color upon ion irradiation from initial white to yellow, light blue, deep blue, light green, and cyan shows an excellent and reversible chromatic response of tungsten oxide nanorods to irradiation. Such a property can be exploited to fabricate radiation sensors. The fragmentation and joining at different energy scales have essential implications in nanodevice fabrication through the bottom-up approach as well as for the development of fusion reactors.