Commercial Fusion Power Plant Research Articles

An integrated digital framework for the design, build and operation of fusion power plants.

The development of a commercial fusion power plant presents a unique set of challenges associated with the complexity of the systems, the integration of novel technologies, the likely diversity and distribution of the organizations involved, and the scale of resources required. These challenges are reviewed and compared to those for other complex engineering systems. A framework for creating a digital environment that integrates research, test, design and operational data is discussed and is based on combining the integrated nuclear digital environment (INDE), proposed recently for nuclear fission power plants, with the hierarchical pyramid of test and simulation used in the aerospace industry. The framework offers the opportunity to plan modelling strategies that allow large design domains to be explored prior to optimizing a detailed design for construction; and in this context, the relationship between measurements and predictions are explored. The use of the framework to guide the socio-technical activity associated with a distributed and collaborative design process is discussed together with its potential benefits and the technology gaps that need to be addressed in order to realize them. These benefits include shorter development times, reduced costs and improvements in credibility, operability, reliability and safety.

Towards the EU fusion-oriented neutron source: The preliminary engineering design of IFMIF-DONES

The need of a high-intensity, 14 MeV neutron source for the qualification of materials under fusion-relevant conditions has been recognized in the European fusion programme as an essential step towards the design and licensing of DEMO and future commercial fusion power plants. This need has pushed the EU to support the development of a Li(d,nx) neutron source called IFMIF-DONES (International Fusion Materials Irradiation Facility-DEMO Oriented NEutron Source) based on and taking advantage of the results obtained in the IFMIF/EVEDA (IFMIF Engineering Validation and Engineering Design Activities) project conducted in the framework of the bilateral EU-Japan Broader Approach Agreement. The design activities and the supporting R&D work of the IFMIF-DONES facility are presently being carried out in the framework of the Work Package Early Neutron Source (WPENS) of the EUROfusion Consortium in direct collaboration with F4E Agency, with the main goal of consolidating the underlying technology and developing a sound design basis in order to be ready for IFMIF-DONES construction at the early beginning of the next decade.In this paper, some important aspects of the IFMIF-DONES Preliminary Engineering Design concerning in particular the main design integration issues in the target area are presented and briefly discussed.

Thermal-hydraulic analysis of the First Wall of a CO2 cooled pebble bed breeding blanket for the EU-DEMO

The Demonstration Fusion Reactor in Europe (EU-DEMO) is considered to be the last step before building a commercial fusion power plant. Up to now, He has been chosen as gas coolant for three of the four candidate breeding blankets proposed for the EU-DEMO, being the Helium Cooled Pebble Bed (HCPB) among them. The choice of He coolant is supported, among other features, by its superior heat transfer capabilities, its transparency to neutrons and chemical inertness. However, the use of this coolant in DEMO poses some feasibility questions, like the large coolant inventory, resulting in possible safety issues, as well as the large circulating power. On the other side, CO2 has been the fluid coolant choice in the nuclear fission industry for gas-cooled reactors since the 1950s. The key advantage of this gas is its larger density, which results in more compact components, lower coolant inventories, lower circulating powers and thus better plant thermal efficiency. This, coupled also with a relatively high molecular stability at moderate temperatures (up to 600 °C), its larger availability and lower price make CO2 an attractive coolant for a pebble bed breeding blanket, as an alternative to He.As a Plasma Facing Component, the First wall (FW) is one of the components under the most challenging environment. Currently, the FW region at the upper port inboard location is expected to suffer the most critical heat flux loads. Indeed, a peak charged particle heat flux 0.99 MW/m2 together with a radiation heat flux of 0.18 MW/m2 is foreseen at this location. This paper reports the thermal-hydraulic analyses carried out to find the practical heat flux limits by using CO2 as coolant. The analyses have been performed by using the commercial computational fluid dynamics (CFD) code ANSYS CFX. Different augmented technical wall roughness areas have been considered, for a best heat transfer-to-pressure drop ratio. Sensitivity analyses have been performed as well for higher peak particle heat fluxes, evidencing that a CO2-cooled FW can dissipate peak heat fluxes up to 1.8 MW/m2 keeping the circulator power in a reasonable range.

Comparative analysis of the efficiency of a CO2-cooled and a He-cooled pebble bed breeding blanket for the EU DEMO fusion reactor

The EU DEMO is considered to be the last step before building a commercial fusion power plant. Up to now the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) is one of the four BB concepts being investigated in the EU for their possible implementation in the EU DEMO.While He is the reference gas cooling for the BBs in DEMO, CO2 has been the coolant choice in the nuclear fission industry for gas-cooled reactors during the past 60 years. In the recent years, it is also gaining interest as a secondary coolant for the Balance of Plant (BoP) to replace the traditional vapor (Rankine) cycle. Despite the lower heat capacity and thermal conductivity of CO2 with respect to He, the larger density of the former can lead to an advantageous performance on heat dissipation rate per unit of pumping power. This feature, coupled with its relative transparency to neutrons, the molecular stability at the operating temperatures of the BB and the industrial scale ready turbomachinery makes CO2 an attractive alternative to He.The aim of this paper is to make a parametric study of the gross and net efficiencies and the design of a simplified BoP of a DEMO power plant based on a CO2-cooled compared to the reference EU DEMO HCPB BB, as a function of the BB outlet temperature and turbine inlet pressure. From this study it is concluded that the vacuum vessel and the divertor heat sources are benefitial in order to improve the net plant efficiency for both coolants. Also, it has been observed that the plant net efficiency is better for the case of a PHTS with CO2 as heat transfer fluid under the assumption that both BBs share the same geometry. In the case that the BBs do not share the same geometry, the study concludes that a PHTS with CO2 leaves a pressure drop margin of about 1.5 before reaching the same net efficiency as a HCPB BoP DEMO plant.

Thermo-mechanical stress analysis of the water cooled DEMO First Wall mock-up components

The European DEMOnstration (DEMO) reactor precedes the development of the fully commercial fusion power plant. One of the key components of the DEMO is the first wall (FW). The FW is acted upon by extreme particle and radiation loads, which results in high heat fluxes varying across the reactor vessel. This paper attempts to evaluate two DEMO FW water cooling concepts from the stress point of view. The heat sink material chosen for the FW is Eurofer97 with thermal conductivity of ∼28 W/mK at 20 °C; the cooling conditions opted are those of pressurized water reactor (PWR), i.e. 285 °C and 15.5 MPa.In the present work it is attempted to perform the stress analysis of the two different heat sink channel configurations, these are circular pipes with countercurrent flow and channel with hypervapotron configuration. The thermo-hydraulic optimization of these components has been performed in one of the author’s previous work. The stress in the components is caused by the pressure of the fluid and increase in the temperature of the material. To accomplish the work, commercial finite element method software ANSYS is used. In a case of unsatisfactory results, further recommendations are given. Simulations are performed with P91, which is material being commonly used in power engineering and will be also used for manufacturing of the discussed configuration of DEMO FW mock-ups which will be afterwards tested in high heat flux facility HELCZA.

Plasma Control Requirements for Commercial Fusion Power Plants: A Quantitative Scenario Analysis With a Dynamic Fusion Power Plant Model

The authors constructed a dynamic simulation model of a nuclear fusion power plant on Modelica language to obtain fundamental knowledge on the plasma control requirements for the future commercial fusion power plants. The fusion power plant model was designed with a 1500-MW thermal output tokamak reactor with He-cooled Li2TiO3 solid breeder blanket (coolant outlet conditions: 8 MPa and 515.8 °C). A superheated Rankine cycle was designed to achieve the electrical output of 485.38 MW with the operating pressure of 20.5 MPa. Two plasma output patterns, a step decrease of power and a single pulse decrease of power, were simulated to assess the response of the power plant. A sudden step decrease in fusion neutron led to an immediate decrease in the blanket temperature and the first coolant temperature. In order to avoid the sharp temperature drop, a need for a turbine bypass mechanism or a He coolant boiler bypass mechanism was indicated. On the other hand, because of the delay in the plant responses, the deviation of the electrical output from steady state could be minimized by recovering the plasma output in few tens of seconds. Based on the findings, a new diagram was presented that illustrates an important plasma control requirements for future commercial fusion power plants.

Magnetic confinement fusion: a brief review

Fusion energy is considered to be the ultimate energy source, which does not contribute to climate change compared with conventional fossil fuel. It is massive compared with unconventional renewable energy and demonstrates fewer safety features compared with unconventional fission energy. During the past several decades, never-ceasing efforts have been made to peacefully utilize the fusion energy in various approaches, especially inertial confinement and magnetic confinement. In this paper, the main developments of magnetic confinement fusion with emphasis on confinement systems as well as challenges of materials related to superconducting magnet and plasmafacing components are reviewed. The scientific feasibility of magnetic confinement fusion has been demonstrated in JET, TFTR, JT-60, and EAST, which instigates the construction of the International Thermonuclear Experimental Reactor (ITER). A fusion roadmap to DEMO and commercial fusion power plant has been established and steady progresses have been made to achieve the ultimate energy source.

Examples of What Fusion Energy Can Contribute to the Near Term Needs of Society Before We Build the First Fusion Power Plant

It will be several decades before the first commercial fusion power plant is placed on the electricity grid. However, there are many non-electricity uses for fusion energy that can be realized in the next 5–10 years when the Q value (energy out/energy in) obtained in fusion systems is ≪1. This paper explores those possibilities dividing them into 3 categories: 1) Commercial Products, 2) Applications for Academia and Utilities, 3) Applications for Military/Governmental uses. At the present time, over 20 near term applications have been identified but it is expected that there could be many more once the fusion community focuses on the near term time frame. Since the near term applications do not require Q > 1, all fusion reactions can be considered. This means, for example, that one can consider fusion reactions that not only emit neutrons (from DT and DD) but also other reactions such as D3He that emit high-energy (14 MeV) protons because they can be used to make short half life Positron Emission Tomography (PET) isotopes. Both steady state and pulsed fusion reactions are also fair game because there is no reason for Q > 1 to be a constraint.

From W7-X to a HELIAS fusion power plant: On engineering considerations for next-step stellarator devices

With the recent start of operation of Wendelstein 7-X (W7-X), the helical-axis advanced stellarator concept (HELIAS) has gained renewed attention. In particular a discussion has been started about a research strategy leading from W7-X to a commercial HELIAS fusion power plant. In order to bridge the respective gap in physics and technology between those devices, concepts for an intermediate-step burning plasma stellarator are under discussion. Recent studies have focused on physics issues of importance to such devices. Extending this discussion, engineering and technology considerations for next-step HELIAS devices are discussed in this work.

Soldered joints—an essential component of demountable high temperature superconducting fusion magnets

Demountable superconducting magnet coils would offer significant benefits to commercial nuclear fusion power plants. Whether large pressed joints or large soldered joints provide the solution for demountable fusion magnets, a critical component or building block for both will be the many, smaller-scale joints that enable the supercurrent to leave the superconducting layer, cross the superconducting tape and pass into the solder that lies between the tape and the conductor that eventually provides one of the demountable surfaces. This paper considers the electrical and thermal properties of this essential component part of demountable high temperature superconducting (HTS) joints by considering the fabrication and properties of jointed HTSs consisting of a thin layer of solder (In52Sn48 or Pb38Sn62) sandwiched between two rare-earth-Ba2Cu3O7 (REBCO) second generation HTS coated conductors (CCs). The HTS joints are analysed using numerical modelling, critical current and resistivity measurements on the joints from 300 to 4.2 K in applied magnetic fields up to 12 T, as well as scanning electron microscopy studies. Our results show that the copper/silver layers significantly reduce the heating in the joints to less than a few hundred mK. When the REBCO alone is superconducting, the joint resistivity (RJ) predominantly has two sources, the solder layer and an interfacial resistivity at the REBCO/silver interface (∼25 nΩ cm2) in the as-supplied CCs which together have a very weak magnetoresistance in fields up to 12 T. We achieved excellent reproducibility in the RJ of the In52Sn48 soldered joints of better than 10% at temperatures below Tc of the REBCO layer which can be compared to variations of more than two orders of magnitude in the literature. We also show that demountable joints in fusion energy magnets are viable and need only add a few percent to the total cryogenic cost for a fusion tokamak.

Present status of liquid metal research for a fusion reactor

Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine.Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far.In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

Considerations on the DEMO pellet fuelling system

The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

Current Status of Beryllium Materials for Fusion Blanket Applications

Beryllium is a promising functional material for several breeder system concepts to be tested within the experimental fusion reactor ITER and, later, implemented in the first commercial demonstration fusion power plant DEMO. For these applications its resistance to neutron irradiation and the detrimental effects of radiogenic gases (helium and tritium) is crucial for fusion reactor safety, subsequent waste management and material recycling. A reliable prediction of beryllium behavior under fusion irradiation conditions requires both dedicated experiments and advanced modeling.Characterization of the reference and alternative beryllium pebble grades was performed in terms of their microstructure and tritium release properties. The results are discussed with respect to their application in fusion blanket systems.The outcomes from the HIDOBE-01 post irradiation experiment (PIE) are discussed to highlight several interesting features manifested by beryllium irradiation at fusion relevant temperatures.Titanium beryllide is presently developed as a possible substitute for beryllium pebbles as it shows better oxidation resistance, higher melting temperature and tritium release efficiency. Pebbles consisting predominantly of Be12Ti phase were successfully fabricated at Rokkasho, Japan.Recent advances in modeling provide new insights on the production of point defects and the behavior of helium and hydrogen impurities in beryllium, improving understanding of the mechanisms of primary damage production, hydrogen’s effect on the size and the shape of gas bubbles, and tritium removal from the pebbles.The relevance of the experimental and modeling results on irradiated beryllium for the design of a fusion demonstration reactor is evaluated, and recommendations for future R&D programs are proposed.

Summary of the International Workshop on Magnetic Fusion Energy (MFE) Roadmapping in the ITER Era; 7–10 September 2011, Princeton, NJ, USA

With the ITER project now well under way, the countries engaged in fusion research are planning, with renewed intensity, the research and major facilities needed to develop the science and technology for harnessing fusion energy. The Workshop on MFE Roadmapping in the ITER Era was organized to provide a timely forum for an international exchange of technical information and strategic perspectives on how best to tackle the remaining challenges leading to a magnetic fusion DEMO, a nuclear fusion device or devices with a level of physics and technology integration necessary to cover the essential elements of a commercial fusion power plant. Presentations addressed issues under four topics: (1) Perspectives on DEMO and the roadmap to DEMO; (2) Technology; (3) Physics-Technology integration and optimization; and (4) Major facilities on the path to DEMO. Participants identified a set of technical issues of high strategic importance, where the development strategy strongly influences the overall roadmap, and where there are divergent understandings in the world community, namely (1) the assumptions used in fusion design codes, (2) the strategy for fusion materials development, (3) the strategy for blanket development, (4) the strategy for plasma exhaust solution development and (5) the requirements and state of readiness for next-step facility options. It was concluded that there is a need to continue and to focus the international discussion concerning the scientific and technical issues that determine the fusion roadmap, and it was suggested that an international activity be organized under appropriate auspices to foster international cooperation on these issues.

Safety Analyses for CANDOR, an Advanced-Fuel Fusion Device

In a Deuterium-Tritium fusion reactor, nearly 20% of the thermal power has to be transferred from the hot plasma through the wall components of the burn chamber. Design requirements of commercial fusion power plant in-vessel components are potentially even more stringent than those of experimental devices. Fusion nuclear reactor studies are currently devoted mostly to the Deuterium-Tritium (DT) fuel cycle, since it is the easiest way to reach ignition or a high energy gain. However, reducing the activation of materials is one of the biggest concerns for fusion power: the study of advanced fuel fusion devices, such as the CANDOR Deuterium-Helium-3 (DHe3) tokamak, is proposed for this purpose. The plasma confinement requirements for a DHe3 reactor are much more challenging than those for a DT reactor. Thus, the demands on the divertor and the first wall are more severe, particularly during a disruption. Safety analyses, starting from heat load determinations, have been performed for CANDOR, a proposed DHe3 experiment, starting from similar evaluations carried out for the ARIES III DHe3 reactor.

LIFE: The Case for Early Commercialization of Fusion Energy

This paper presents the case for early commercialization of laser inertial fusion energy (LIFE). Results taken from systems modeling of the US electrical generating enterprise quantify the benefits of fusion energy in terms of carbon emission, nuclear waste and plutonium production avoidance. Sensitivity of benefits-gained to timing of market-entry is presented. These results show the importance of achieving market entry in the 2030 time frame. Economic modeling results show that fusion energy can be competitive with other low-carbon energy sources. The paper concludes with a description of the LIFE commercialization path. It proposes constructing a demonstration facility capable of continuous fusion operations within 10 to 15 years. This facility will qualify the processes and materials needed for a commercial fusion power plant.

Investigation of Lead Target Nuclei Used on Accelerator-Driven Systems for Tritium Production

High-current proton accelerators are being researched at Los Alamos National Laboratory and other laboratories for accelerator production of tritium, transmuting long-lived radioactive waste into shorter-lived products, converting excess plutonium, and producing energy. These technologies make use of spallation neutrons produced in (p,xn) and (n,xn) nuclear reactions on high-Z targets. Through (p,xn) and (n,xn) nuclear reactions, neutrons are produced and are moderated by heavy water. These moderated neutrons are subsequently captured on 3He to produce tritium via the (n,p) reaction. Tritium self-sufficiency must be maintained for a commercial fusion power plant. Rubbia succeeded in a proposal of a full scale demonstration plant of the Energy Amplifier. This plant is to be known the accelerator-driven system (ADS). The ADS can be used for production of neutrons in spallation neutron source and they can act as an intense neutron source in accelerator-driven subcritical reactors, capable of incinerating nuclear waste and of producing energy. Thorium and Uranium are nuclear fuels and Lead, Bismuth, Tungsten are the target nuclei in these reactor systems. The spallation targets can be Pb, Bi, W, etc. isotopes and these target material can be liquid or solid. Naturally Lead includes the 204Pb (%1.42), 206Pb (%24.1), 207Pb (%22.1) and 208Pb (%52.3) isotopes. The design of ADS systems and also a fusion-fission hybrid reactor systems require the knowledge of a wide range of better data. In this study, by using Hartree–Fock method with an effective nucleon-nucleon Skyrme interactions rms nuclear charge radii, rms nuclear mass radii, rms nuclear proton, neutron radii and neutron skin thickness were calculated for the 204, 206, 208Pb isotopes. The calculated results have been compared with those of the compiled experimental and theoretical values of other studies.

Near Term Applications of Inertial Electrostatic Confinement Fusion Research

For the past 15 years, the Inertial Electrostatic Confinement (IEC) fusion group at the University of Wisconsin-Madison has been conducting experiments to demonstrate that there can be many near term applications of fusion research long before the production of electricity in commercial fusion power plants. This research has concentrated on three fuel cycles: DD, D3He, and 3He3He. Some of the major accomplishments are listed below:a. The production of > 108 DD neutrons per second on a steady state basisb. The production of pulsed DD neutrons to over 1010 per second in 10Hz, 100 μs bursts.c. The production of 14.7 MeV protons at > 108 per second (steady state) from the D3He reaction.d. Demonstrated the detection of the explosive C-4 with steady state DD neutrons.e. Demonstrated the detection of Highly Enriched U (HEU) with pulsed DD neutron fluxes.f. Production of the positron emission tomography (PET) isotopes, 94mTc and 13Nusing D3He protons.g. Production of the first measured 3He3He fusion reactions in an IEC device.h. Development of unique diagnostic techniques to measure the rate, spectrum, and location of fusion reactions in IEC devices.i. Use of an IEC device to study the behavior of materials at high temperature during charged particle bombardment.The accomplishments above were carried out in 3 devices HOMER, 3HeCTRE, and HELIOS that have operated up to 180 kV and meter currents of 65 mA. New applications are currently being explored and expanded roles for the IEC device will be described.

DEMO development strategy based on China FPP program

The DEMO in China is aimed to demonstrate the safety, reliability and environmental feasibility of the fusion power plants (FPP), meanwhile to demonstrate the feasibility of the prospective economic evaluation of commercial fusion power plants. As known, there is still a long way to go towards an economically competitive commercial power plant, hence DEMO in China is an indispensable step. There are two options of breeding blankets with ceramic and lithium-lead breeders as DEMO blanket concepts under the conditions of meeting the requirement of the neutronics, thermal-hydraulics and mechanics aspects. In this paper, the DEMO blanket development strategy, conceptual design, and related R&D activities, based on China fusion power plants program are presented.

Power plant conceptual studies in Europe

The European fusion programme is ‘reactor oriented’ and it is aimed at the successive demonstration of the scientific, the technological and the economic feasibility of fusion power. The European Power Plant Conceptual Study (PPCS) has been a study of conceptual designs of five commercial fusion power plants and the main emphasis was on system integration. It focused on five power plant models which are illustrative of a wider spectrum of possibilities. They are all based on the tokamak concept and they have approximately the same net electrical power output, 1500 MWe. These span a range from relatively near-term, based on limited technology and plasma physics extrapolations, to an advanced conception.The PPCS allows one to clarify the concept of DEMO, the device that will bridge the gap between ITER and the first-of-a-kind fusion power plant. An assessment of the PPCS models with limited extrapolations highlighted a number of issues that must be addressed to establish the DEMO physics and technological basis.