Commercial Fusion Power Research Articles

Plasma control for the step prototype power plant

In 2019 the UK launched the Spherical Tokamak for Energy Production (STEP) programme to design and build a prototype electricity producing nuclear fusion power plant, aiming to start operation around 2040. The plant should lay the foundation for the development of commercial nuclear fusion power plants. The design is based on the spherical tokamak principle, which opens a route to high pressure, steady state, operation. While facilitating steady state operation, the spherical design introduces some specific plasma control challenges: (i) All plasma current during the burn phase should to be generated through non-inductive means, dominated by bootstrap current. This leads to operation at high normalised plasma pressure βN with high plasma elongation, which in turn imposes effective active stabilisation of the vertical plasma position. (ii) The tight aspect ratio means very limited space for a central solenoid, imposing that even the current ramp up must be non-inductively generated. (iii) The compact design leads to extreme heat loads on plasma facing components. A double null design has been chosen to spread this load, putting strict demands on the control of the unstable vertical plasma position. (iv) The heat pulses associated with unmitigated ELMs are unlikely to be acceptable imposing ELM free operation or active ELM control. (v) To reduce and spread heat loads, core and divertor radiation and momentum loss has to be controlled, aiming to operate with simultaneously detached upper and lower divertors. (vi) High pressure operation is likely to require active resistive wall mode (RWM) stabilisation. (vii) The conductivity distribution in structures near the plasma must be carefully selected to reduce the growth rates for the vertical instability and the RWM without damping the penetration of the of magnetic fields from active control coils too much. This article describes the initial work carried out to develop a STEP plasma control system.

Optimisation of the poloidal field system for advanced divertor configurations in STEP

The power exhaust proves to be one of the most challenging and concept–defining aspects in the design of a commercial fusion power plant, while the magnetic coil system, capable of supporting advanced exhaust solutions, emerges as one of the main design and cost drivers. Consequently, much effort should be dedicated to the optimisation of a robust global magnetic configuration, which integrates both the plasma and edge scenarios, while ensuring engineering feasibility and compatibility with the available technology. Here we present a multidisciplinary framework employed to analyse, evaluate, and optimise the Spherical Tokamak for Energy Production (STEP) equilibrium configuration, coupled with a viable divertor solution, and a compatible poloidal field coil system. The complexity of this task leads to a multitude of potentially conflicting requirements and competing constraints. We identify interfaces and conflicts between the aspects of the design that were previously considered independently, and highlight the key benefits, trends, and trade–offs between alternative configurations. We demonstrate that advanced exhaust solutions, simultaneously applied to both inboard and outboard divertors, are accessible with feasible coil sets under conditions relevant for STEP. We show that the most promising inner–X geometry, paired with the outer super–X configuration, can significantly enhance divertor’s power handling capability, allowing access to stable detached regimes. The coil set feasibility is further assessed considering its compatibility with the assumed plasma initiation scenario, and with the most demanding plasma current density profiles utilising alternative heating and current drive schemes.

Approach to startup inventory for viable commercial fusion power plant

With the increasing efforts to commercialize fusion power, private and government organizations are investing heavily in the development of technology to support a viable fusion power plant. Deuterium-Tritium (DT) fueled reactors are more prevalent than other proposed designs, requiring tritium processing and handling technology for safe operations and self-sufficiency. Further, each fusion power plant will need a specific-to-design startup inventory of tritium to begin operations. This startup inventory is required prior to breeding and is the minimum tritium inventory required to fill each processing component in the fuel cycle, to offset radioactive decay losses, and to avoid a zero-fuel situation for continuous operation. We present an approach to calculate the startup tritium inventory for a 500 MWth reactor, with considerations for reserve inventory for maintenance and commissioning. A baseline startup inventory was calculated to be approximately 327 gs. This value was obtained using modest assumptions about the technology and operating parameters of a fusion power plant. The required operating reserve inventory or the inventory necessary to keep a fusion power plant operational using only direct internal recycling for 24 h for the same plant design is approximately 642 gs. The approach and findings of this paper will enable fusion energy stakeholders to better utilize the existing scarce global tritium supply.

Legal Issues of Fusion Technologies

Abstract A comprehensive regulatory framework for fusion must consider the unique characteristics and challenges of fusion technology, including safety, environmental protection, waste management, export/import control, and nuclear liability. The International Atomic Energy Agency (IAEA) has established safety standards and guidelines related to fusion, but more needs to be done to address the gaps in the existing legal frameworks. The article emphasizes the need for clear and predictable regulations to provide confidence for investors and promote the research, development, and deployment of this promising technology. Although significant progress has been made in recent years, commercial fusion power is still likely many years away from being a reality, but a legal framework is essential for its eventual success.

Updates of the removable biological shielding blocks inside of the test cell, part of the test systems of IFMIF-DONES

The International Fusion Materials Irradiation Facility - DEMO Oriented Neutron Source (IFMIF-DONES) is a facility which is designed under the framework of the EU fusion roadmap. It is going to be an essential irradiation facility for testing and qualifying candidate materials under severe irradiation conditions of a neutron field having an neutron irradiation effect on materials like the one expected in a commercial fusion power reactor. The material specimens are irradiated in a containment structure named Test Cell (TC), which is part of the Test Systems (TS).The first protecting „barrier” against irradiation affecting the surrounding of the TC, are the Removable Biological Shielding Blocks (RBSBs). These ∼9 m height, ∼80 tons elements, which are formed by stainless steel liners filled with heavy concrete, need to be remotely handled as after the first experiments they will get dose rates above the hands-on limit. Irradiation will also result in a large nuclear heat power deposited in the shielding blocks, therefore needs to be actively water cooled by a system of embedded pipes to control the temperature.In this paper the updated design of the RBSBs is described, including the latest achievements and proposals for feasible manufacturing, lifting and positioning possibilities of the blocks inside of the TC respecting the given tolerances.

Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities

The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021.

Helium production and material damage rate assessment in EU DEMO HCPB divertor

EU DEMO should be the first demonstrative fusion power plant, and it is expected to show that net electric energy can be produced and supplied to the grid. The availability and lifetime of the plant will be comparable to that of a commercial fusion power plant. The capacity to provide considerable net electric power safely and continuously to the grid (from 100 MW to 500 MW) is one of DEMO's significant purposes. The neutrons cause defects, leading to the transmutation of elemental atoms when interacting with the divertor component materials. These reactions produce gases, notably helium, which can induce material swelling and embrittlement. In this paper, the DPA and helium production in divertor has been assessed, mainly focusing on the structural material: EUROFER. Helium production rates reach around 117 appm/fpy as the peak, while DPA values reach about 4 dpa/fpy as the maximum value. These calculations were performed using the MCNP6 code with FENDL 3.2 nuclear data and the ADVANTG tool.

The IAEA DEMO Programme Workshop Series: 2012–2021 report

The IAEA DEMO Programme Workshop Series was initiated to be the reference forum to debate, evaluate and establish the next steps to be taken within the international fusion community to deliver fusion as a reliable source of clean energy. The DEMO programme refers predominantly to a future magnetic confinement Tokamak design concept, after the ITER project, with a goal of delivering electricity to the grid. To enable this, the programme workshop series provides a unique frame where the discussion and analysis of the progress and findings of the various DEMO programmes, and not just the presentation of results being the major goals. More recently, due to the construction development of ITER, the workshop has also provided a forum to bring together the fusion community and industry. This is an important development for DEMO programmes, which will be more reliant on industry given their focus on devices proximate to a commercial fusion power plant. In this paper a summary of all editions of this workshop (2012–2013–2015–2016–2018–2019) up to 2021 are summarized. Topics addressed are listed and findings and open questions pointed out for each edition.

The effect of in situ irradiation on the superconducting performance of REBa2Cu3O7−δ-coated conductors

Commercial fusion power plants will require strong magnetic fields that can only be achieved using state-of-the-art high-temperature superconductors in the form of REBa2Cu3O7−δ-coated conductors. In operation in a fusion machine, the magnet windings will be exposed to fast neutrons that are known to adversely affect the superconducting properties of REBa2Cu3O7−δ compounds. However, very little is known about how these materials will perform when they are irradiated at cryogenic temperatures. Here, we use a bespoke in situ test rig to show that helium ion irradiation produces a similar degradation in properties regardless of temperature, but room-temperature annealing leads to substantial recovery in the properties of cold-irradiated samples. We also report the first attempt at measuring the superconducting properties while the ion beam is incident on the sample, showing that the current that the superconductor can sustain is reduced by a factor of three when the beam is on.Impact statementREBa2Cu3O7−δ high-temperature superconductors are an enabling technology for plasma confinement magnets in compact commercial fusion power plants, owing to their ability to carry very high current densities when processed as quasi-single crystals in the form of coated conductors. In service in a fusion device, the magnet windings will be exposed to a flux of fast neutrons that will induce structural damage that will adversely affect the superconducting performance, but very little data are currently available on the effect of irradiation at the cryogenic temperatures relevant for superconducting magnets. Moreover, even room-temperature annealing substantially affects superconducting properties after irradiation, so to obtain key technical data for fusion magnet designers, it is important to measure these properties in situ, under irradiation. This work shows that for the first time, it is important to consider how energetic particles directly influence superconductivity during irradiation because we observe a reduction in zero-resistance current by a factor of as much as three when an ion beam is incident on the sample. Although neutrons will not interact with the material in the same way as charged ions, primary knock-on ions from neutron damage are expected to have a similar effect to the He+ ions used in our study.Graphical abstract

Radiation Damage Analysis of FNSF Components Using McCad and MCNP

The Fusion Energy System Studies Fusion Nuclear Science Facility (FESS-FNSF) concept represents a transitional step between ITER and a commercial fusion power plant. The FNSF is a conceptualized D-T fueled tokamak with 518 MW of fusion power that has been extensively used to explore and optimize design features. The energetic 14.1-MeV neutrons can produce significant localized heating and activations, and can cause damage to plasma-facing components, which can determine maintenance/outage scheduling needs and also impact the lifetime of the device as a whole. This study illustrates a neutronics analysis that was conducted on a 22.5-degree symmetric sector of the FNSF with the goal of understanding the neutron heating and radiation damage that can be characterized by quantifying the displacements per atom (dpa). Concurrently, this study also focused on the development of analysis capabilities by converting a three-dimensional computer-aided design model of the FNSF into MCNP6.2 input using the McCad code. Accordingly, some confirmatory results on tritium production and the tritium breeding ratio (TBR) are provided to support model validation. The results produced by MCNP6.2 simulations showed that the highest heating and damage occurred in the outboard region, which concentrated approximately 290 MW of the total nuclear heating, in contrast to 97 MW within the inboard region. These results are consistent with previous studies that employed earlier versions of the FNSF concept and different modeling approaches. This study also provides additional details on neutron wall loading, as well as total heating from neutrons and gammas, results which show the total heating of the device (16 sectors) is approximately 477.83 ± 0.80% MW, indicating a neutron energy multiplication factor of 1.15. Additionally, the capability to calculate hydrogen and helium production, as well as dpa, is illustrated. Finally, the neutronics effects of using alternative materials to tungsten carbide were evaluated for the vacuum vessel, low-temperature shield, and structural ring components, which showed that compounds like YH2, Mg(BH4)2, and ZrH2 could reduce the total heating on the magnet and also reduce the TBR.

Technological features of a commercial fusion power plant, and the gap from DEMO

Technological features of a commercial fusion power plant, and the gap from DEMO

Preliminary Development and Verification of Divertor Module for CFETR System Safety Analysis Code

China Fusion Engineering Test Reactor (CFETR), the next-generation fusion device of China, is designed to bridge the gap between International Thermonuclear Experimental Reactor (ITER) and the commercial fusion power plant by solving the essential physical, engineering, and technical problems, including tritium self-sufficiency and high duty cycle of 0.3–0.5. Before the completion of the design and construction of CFETR, the thermal-hydraulic safety assessment is needed in order to ensure the safety of nuclear facilities and to get the national nuclear safety license. Up to now, there is no specialized code for the fusion reactor safety assessment. And the safety assessment work that had been done for the fusion device is completed by the system safety analysis code for pressurized water reactors, such as RELAP5, CHATHARE, and so on. For developing a simulation tool that can meet the safety assessment needs of the fusion reactor, some special models need to be established compared to the fission reactor. In this article, the development of a module for the divertor is presented. The thermal-hydraulic safety of the divertor with design basis accidents, and in-vessel loss of coolant (LOCA), is analyzed based on the developed code. Comparisons are performed to verify the reliability and feasibility of the developed code.

Research on Nondestructive Examination of Jacket Section for CFETR TF Coil

The Chinese Fusion Engineering Testing Reactor (CFETR) is aiming to bridge the gap between the International Thermonuclear Experimental Reactor (ITER) and the first commercial fusion power plant, a necessary complement of the ITER. The requirement for sustaining long burn duration as specified in the duty time CFETR mission necessitates the use of superconducting magnets. Totally 16 toroidal field (TF)D-shaped coils which are of cable-in-conduit conductor type are designed. All the subcomponents have very demanding requirement to withstand the severe environments. This paper gives the nondestructive examination (NDE) method developed for the jacket of TF conductor which is made of 316L stainless steel. Based on the linear elastic fracture mechanics, maximum acceptable defect sizes in the TF jacket material have been defined. Phased array ultrasonic testing(PAUT)method and eddy current test (ECT) method were developed for the circular-in-square jacket inspection. For PAUT method, focus laws were calculated and optimized. For ECT method, specific ECT probe was designed and manufactured. Benchmark experiments were carried out to study the NDE reliability.

Experimental Investigation of Tritium Compatible Stainless Steel Roots Pumps

SummaryPart 1 of this paper describes the experiments performed to qualify the Roots pump Okta 1500 M for tritium applications at ITER. In Part 2, the goal was to define performance limits in normal operation mode. These performance limits led to some suggestions and optimization of operating parameters in order to enhance the performance of the pump. These modifications are mainly meant for introducing the purge gas into mechanical seals. This pump is the first pump equipped with dedicated gas lubricated mechanical seals capable to withstand tritium. The task of the experiments described in the article is to discover the best performance as well as the lowest influence of any kind of purge gases leaking into the process gas for tritium. The experiment was also designed to confirm that the pump mechanical seal is rugged against cross migration of process gas to ensure confinement integrity in operation range. With these suggested modifications, the pump should be even better suited for ITER's specification and commercial fusion power applications.

Fusion: it’s hotting up

James McKenzie applauds recent record investments in commercial fusion power plants, which could help us to create a net-zero economy.

SOLPS-ITER modeling of CFETR advanced divertor with Ar and Ne seeding

The Chinese Fusion Engineering Testing Reactor (CFETR) is a project proposed by the Chinese fusion community to bridge the gap between ITER and a commercial fusion power plant with fusion power up to 1 GW. The mitigation of divertor target heat fluxes for such a powerful machine is a challenging problem, which might appear to be more severe than in ITER. In the present paper, the results of the CFETR advanced divertor optimization by SOLPS-ITER modeling with full drifts and currents activated are presented. Three divertor geometries, which differ by the distance from the X-point to the strike point on the outer target, are considered. Argon (Ar) and neon (Ne) are compared as seeded impurities. It is demonstrated that for all three geometries and for both radiators it is possible to achieve acceptable divertor heat loads (below 5 MW m−2) without notable fuel dilution (Z eff < 2.5). Impurity compression in divertors and pedestal radiation are compared for two gases. Similar core plasma and divertor conditions, as well as radiated power fraction, may be achieved with 2–3 times less Ar seeding rate than the Ne one. Estimated radiation from the confined region appears to be small compared to the exhaust power. However, in all modeling cases the T e at the far scrape-off layer part of both targets remains significantly above 5 eV, which might cause tungsten (W) sputtering. Further optimization of target shape will be performed to reduce the electron and ion temperature.

Plasma Scenarios for the DTT Tokamak with Optimized Poloidal Field Coil Current Waveforms

In the field of nuclear fusion, the power exhaust problem is still an open issue and represents one of the biggest problems for the realization of a commercial fusion power plant. According to the “European Fusion Roadmap”, a dedicated facility able to investigate possible solutions to heat exhaust is mandatory. For this purpose, the mission of the Divertor Tokamak Test (DTT) tokamak is the study of different solutions for the divertor. This paper presents the plasma scenarios for standard and alternative configurations in DTT. The Single Null scenario is described in detail. The alternative configurations are also presented, showing the good flexibility of the machine.

An exploratory study on application of big science business ecosystem for K-DEMO project

The goal of this research is to study building of business ecosystem in order to drive K-DEMO R&D, design and construction activities. Fusion energy development in Korea proceeds largely in 4 phases: KSTAR, ITER, K-DEMO and commercial fusion power plant. Particularly, based on that the goal of the K-DEMO project is to demonstrate fusion-driven electricity production and build economic foundation for fusion energy, this study analyzed the expert interviews and industry perception surveys with the purpose of seeking the ways to build the business ecosystem for K-DEMO project. We defined the industrial technologies for K-DEMO core components and based on these, conducted surveys on the perception of industries associated with accelerators, nuclear power and space business as well as fusion industries. The results are applied to identify the current technology status of industry and the demands from industry on K-DEMO R&D, design, construction activities and develop the proposal to build the K-DEMO business ecosystem.

Neutronics assessment of EU DEMO alternative divertor configurations

As a demonstration fusion power plant, EU DEMO has to prove the maturity of fusion technology and its viability for electricity production. The central requirements for DEMO rest on its capability to generate significant net electric power to the grid (300 MW to 500 MW) safely and consistently. Plant availability and lifetime will approach that of a commercial fusion power plant. Operating at such regimes presents many complex challenges, of which one is plasma exhaust. To mitigate the risk that the implementation in preceding experimental devices, namely ITER, does not extrapolate to the requirement of DEMO, alternative solutions must be sought. The investigation of alternative divertor configurations was born out of this motive, seeking to resolve a ‘critical’ challenge for the realisation of DEMO. In this paper, we study the neutronics performance of three concepts: Single Null (SN), Super-X (SX) and X-divertor (XD). This is the first time a preliminary analysis of alternative configurations to the SN baseline has been performed. The shielding proposals and design recommendations presented herein should be integrated with other engineering and physics constraints in future iterations of the chosen divertor concept.

Beyond the physics and demonstration of ignition.

Fusion holds the promise of providing growing world energy demand with a carbon-free power source having a universally available fuel source and attractive safety and environmental characteristics. A significant global effort has been underway for over 50 years aimed at the achievement of fusion by inertial confinement. The effort to date has necessarily emphasized understanding the physics of compressing and heating a small amount of fusion fuel to the high densities and temperatures required for ignition and energy gain. Though steady progress has been and is still being made to achieve the required physics understanding and energy gain, those goals have not yet quite been met. It is timely to put progress toward fusion power by inertial confinement into perspective by developing an updated roadmap. Preparing a roadmap from present achievements to the ultimate goal of commercial fusion power requires formally identifying and implementing complementary efforts on a number of fronts. These include the choice, development and demonstration of high repetition rate compression drivers (e.g. lasers) to succeed present day single-pulse sources; design, fabrication and testing of high gain targets (gain of order 100); development of mass production, cost-effective, target fabrication and delivery systems capable of inserting targets into the reaction chamber several times per second, and demonstrating ability to accurately hit and efficiently compress those targets to reliably produce the required fusion yields; design and demonstration of reaction chambers capable of handling energy yields and target debris clearing at the levels required for achieving high power plant reliability with low induced radioactivity. A robust ongoing effort on competitive power plant conceptual design is necessary to guide the implementation of a roadmap, including the timing and level of effort on the 'beyond ignition' demonstrations. This article is part of a discussion meeting issue 'Prospects for high gain inertial fusion energy (part 1)'.