Fusion Power Plant Design Research Articles

The energy output equilibrium scheme with intermediate energy storage for tokamak fusion power plant

Different from the design of nuclear fission power plant, the design of fusion power plant needs to consider the problem of discontinuous fusion energy output from nuclear island. This paper proposed an energy output equilibrium scheme for fusion power plant. Based on analysis and comparison, thermal energy storage method was adopted, and two system operation schemes for fusion power plant were further proposed. Finally, system thermal storage calculation and analysis were conducted under WCCB blanket and HCCB blanket respectively, and conclusions were drawn on the applicable system operation scheme under different blankets. This study provides research foundation and engineering reference for the design and construction of future fusion power plants.

The commoditization of civil nuclear power.

The commoditization of nuclear power through the factory production of sealed micro-power units within a digitally enabled holistic assurance framework is described. This would revolutionize nuclear power-plant design, construction, operation and decommissioning through a paradigm shift to manufacture-operate-remove-recycle (MORR). The potential impact of recent research on an integrated nuclear digital environment for large bespoke nuclear power plants and the design, build and operation of fusion power plants using such a digital environment is explored. These strands are interwoven to discuss the technical, economic and socio-political implications of MORR in the context of micro-reactors and to consider the potential evolution of safeguarding issues based on a digital assurance framework that leads to type approvals. Commoditization of nuclear power would lower costs in line with offshore wind and the output from a single production line in a factory could replace a third of current fossil fuel-based electricity generation in the UK over a 15-year period, making a significant contribution to achieving zero greenhouse gas emissions. The challenges associated with the changes in culture, both in the nuclear industry and in society, as well as the technology gaps, that need to be addressed in realizing this paradigm shift are identified and discussed.

Elevating zero dimensional global scaling predictions to self-consistent theory-based simulations

We have developed an innovative workflow, Stability, Transport, Equilibrium, and Pedestal (STEP)-zero-dimensional (0D), within the OMFIT integrated modeling framework. Through systematic validation against the International Tokamak Physics Activity global H-mode confinement database, we demonstrated that STEP-0D, on average, predicts the energy confinement time with a mean relative error of less than 19%. Moreover, this workflow showed promising potential in predicting plasmas for proposed fusion reactors such as the affordable, robust, compact (ARC) reactor, the European demonstration power plant (EU-DEMO), and the China fusion engineering test reactor (CFETR) indicating moderate H-factors between 0.9 and 1.2. STEP-0D allows theory-based prediction of tokamak scenarios, beginning with 0D quantities. The workflow initiates with the PRO-create module, generating physically consistent plasma profiles and equilibrium using the same 0D quantities as the IPB98(y,2) confinement scaling. This sets the starting point for the STEP module, which further iterates between theory-based physics models of equilibrium, core transport, and pedestal to yield a self-consistent solution. Given these attributes, STEP-0D not only improves the accuracy of predicting plasma performance but also provides a path toward a novel fusion power plant design workflow. When integrated with engineering and costing models within an optimization, this new approach could eliminate the iterative reconciliation between plasma models of varying fidelity. This potential for a more efficient design process underpins STEP-0D's significant contribution to future fusion power plant development.

Nuclear data for fusion: inventory validation successes and future needs

Nuclear data, describing neutron reaction probabilities (cross sections) and decay behaviour, are critical to the design and operation of fusion experiments and future fusion power plants. Equally vital, are the inventory codes that use the data to predict neutron-induced activation and transmutation of materials, which will define the radiological hazards that must be managed during reactor operation and decommissioning. Transmutation, including gas production, combined with the neutron-induced displacement damage, will also cause the properties of materials to degrade, for example through swelling and embrittlement, eventually limiting the lifetime of components. Thus validated and accurate nuclear data and inventory codes are essential. For data validation there are decay heat measurements performed at FNS in Japan more than 20 years ago. The experiments produced an invaluable database for benchmarking of nuclear data libraries; the latest versions of several international libraries perform well against this data during tests with the FISPACT-II inventory code, although there is still scope for improvement. A recent attempt to provide fusion-relevant validation based on γ-spectroscopy data from neutron-irradiated material samples tests produced predictions for short-lived (several hours or less) radionuclides. The detailed analysis performed for molybdenum demonstrates how these data could eventually provide a new benchmark, and also illustrates the potential benefits of further experiments targeting the longer-lived radionuclides relevant to maintenance and decommissioning timescales. There are also some successful tests of transmutation predictions with FISPACT-II. These direct validations of inventory simulations are critical for lifetime predictions and future experiments should learn lessons from the examples described for tungsten, which demonstrate the importance of an accurate description of the neutron spectrum in experiments. More novel experimental techniques are needed to measure helium production in materials such as Fe and C, but the need to validate the nuclear data evaluations used by simulations should motivate future experimental efforts.

CFD Analysis and Optimization of the DEMO WCLL Central Outboard Segment Bottom-Cap Elementary Cell

In the design of magnetic confinement nuclear fusion power plants, the breeding blanket (BB) plays a crucial role since it must fulfil key functions such as tritium breeding, radiation-shielding, and removal of the heat power generated by the plasma. The latter task is achieved by the first wall (FW) and breeding zone (BZ) cooling systems, which in the water-cooled lithium–lead (WCLL) BB employ pressurized water. Different arrangements of BZ coolant conduits have been investigated in the recent past to identify an efficient layout, which could meet the structural materials’ operational temperature constraint and which could provide the optimal coolant outlet temperature. However, most of the computational fluid dynamic (CFD) analyses that have been carried out until now have been focused on the equatorial WCLL elementary cell of the central outboard segment (COB). The aim of this work is to broaden the analysis to other relevant locations in the blanket. An assessment of the design of the cooling system of the COB bottom-cap elementary BZ cell has been identified as a top design priority due to its different geometry and thermal loads. The cooling efficiency of the BZ and FW systems is investigated to assess if the coolant-appropriate design conditions are matched and the temperature distribution in the cell is analyzed to identify the onset of hot spots. Different layouts of the FW systems are proposed and compared in terms of thermal–hydraulic reliability.

Activated Corrosion Products Evaluations for Occupational Dose Mitigation in Nuclear Fusion Facilities

Activated corrosion products generation in primary heat transfer systems of nuclear fusion facilities is a relevant radiological source term for occupation dose assessments. The formation of the Chalk River Undefined Deposit, already well known in nuclear fission power plants, represents a significant safety issue in fusion applications due to the intense high energy neutron fluences (about 14 MeV in Deuterium-Tritium operation). The activated corrosion products formation is a multi-physical problem. The combined synergy of activation, corrosion, dissolution, erosion, ejection, precipitation, and transport phenomena induces the contamination of coolant loop regions located outside the bio-shield, where scheduled worker operation might take place. The following manuscript shows how activated corrosion products are evaluated for the nuclear fusion power plant design under investigation by the Safety and Environment Work Package (WPSAE) of the Eurofusion Consortium (i.e., the European Demonstration power plant, EU-DEMO). The major issues in activated corrosion products estimations are here exposed and the main results for mass and activity inventories are briefly shown for some main Primary Heat Transfer Systems of EU-DEMO.

The physics of turbulence localised to the tokamak divertor volume

Fusion power plant designs based on magnetic confinement, such as the tokamak design, offer a promising route to sustainable fusion power but require robust exhaust solutions capable of tolerating intense heat and particle fluxes from the plasma at the core of the device. Turbulent plasma transport in the region where the interface between the plasma and the materials of the device is handled - called the divertor volume - is poorly understood, yet impacts several key factors ultimately affecting device performance. In this article a comprehensive study of the underlying physics of turbulence in the divertor volume is conducted using data collected in the final experimental campaign of the Mega Ampere Spherical Tokamak device, compared to high fidelity nonlinear simulations. The physics of the turbulence is shown to be strongly dependant on the geometry of the divertor volume - a potentially important result as the community looks to advanced divertor designs with complex geometry for future fusion power plants. These results lay the foundations of a first-principles physics basis for turbulent transport in the tokamak divertor, providing a critical step towards a predictive understanding of tokamak divertor plasma solutions.

A deterministic method for the fast evaluation and optimisation of the 3D neutron wall load for generic stellarator configurations

The neutronic assessment of a fusion power plant design is usually a challenging and time-consuming task involving experts from several disciplines in order to assemble the geometry, source, as well as carry out computationally heavy Monte Carlo transport simulations. In order to overcome this challenge, we present in this work a deterministic method, which combines all these aspects in a single framework that can directly calculate a key neutronics performance indicator, the neutron wall load (NWL), at minimal computational cost for arbitrary stellarator configurations. As our method is based on simple vector and matrix manipulations, a speed on the order of a few CPU-seconds is achieved, which makes it suitable for optimisation frameworks. We demonstrate with a simple optimisation algorithm that it is possible to use our method to generate a first wall with reduced ‘heterogeneity’ in the NWL distribution.

Conceptual design of a water-cooled first wall component for a tokamak machine

In a tokamak, the first wall is a barrier protecting the internal parts of the machine from the intense fluxes of heat and particles coming from the plasma. This contribution presents the critical issues of a water-cooled first wall component based on steel as structural material, that is a relevant case study for the design of future fusion power plants.

Mechanical support concept of the DEMO breeding blanket

• Description of blanket integration concept. • Stress calculation of breeding blanket segments. • Prediction of Breeding blanket deformation. • Identification of custom-machining concept and prediction of achievable tolerances. • Overview of the various integration issues associated with the design of in-vessel components. The DEMO tokamak architecture is based on large vertical breeding blanket (BB) segments that are accessed from a maintenance hall above the tokamak and are vertically replaced through large upper ports of the vacuum vessel (VV). The feasibility of the BB segments mechanical supports is a prerequisite of this vertical segment architecture. Their design directly impacts on the removal kinematics and the remote handling operations required for release and engagement. The supports must withstand large forces acting on the BB in particular due to electromagnetic (EM) loads. At the same time, they must ensure a sufficiently precise positioning of the BB first wall. Their design also takes into account the significant thermal expansion of the blanket segments that are operated at high temperature avoiding excessive support reaction forces. The BB support concept described in this article does not require fasteners or electrical straps to the VV and therefore much reduces the complexity of the BB remote replacement – a valuable characteristic that would make this concept a milestone in meeting one of the goals defined for the DEMO project: to develop a maintainable fusion power plant design [1] . Each blanket segment is individually supported by the VV without any physical contact to the other blankets or in-vessel components. It relies instead on vertical pre-compression inside the VV due to obstructed thermal expansion and radial pre-compression due to the ferromagnetic force acting on the BB material in the toroidal magnetic field. The verification process did not identify show stoppers. Nonetheless, a further evolution of the concept is required including design improvements to mitigate the high stress levels found in the inboard blankets during plasma disruptions. The fact that no excessively high support reaction forces or large BB deflections were found suggests though that the further development of the concept could be successful.

Verification and validation of mHIT code over TMAP for hydrogen isotopes transport studies in fusion-relevant environments

The accurate prediction of tritium inventory and permeation fluxes in the breeding blanket of a D-T fusion reactor is a key aspect for future thermonuclear power plants licensing. Tritium permeation into structural materials could give rise to potential issues concerning the fuel self-sufficiency and can be lost into the environment with resulting radiological risks for the population. In the frame of hydrogen isotopes transport modelling, the Tritium Migration Analysis Program (TMAP) code is considered a referred code for the safety design of nuclear fusion power plants. It is mainly applied to Plasma Facing Components (PFCs), which are subjected to intense particles implantation and surface heat fluxes.Due to some code limitations and to the need of having a more easy-to-use computer code in a multiphysics framework, in the last years the commercial software COMSOL Multiphysics has been used to assess tritium transport studies as an application to ITER, DEMO and CFETR fusion reactors, but it was never validated against TMAP. Within this paper, a COMSOL-based code named mHIT (multi-trapping Hydrogen Isotopes Transport code) is presented. A set of verification and validation (V&V) problems were addressed, with the aim of substantiating the capabilities of the code in common fusion-relevant experimental set-ups and to include different physics according to the level of detail needed for a given application.

Numerical investigation of the tritium permeation phenomenon through cooling plates in breeding blankets

Understanding and predicting tritium transport is considered a key issue for the design of safe and self-sufficient fusion power plants. Indeed, tritium generated in the breeder materials is susceptible to permeate to the blanket coolant circuits where it is harder to remove and process. Therefore, the predictive models become very important to evaluate the performance of any blanket design. In this work, 2D finite volume methods are employed to analyze in detail the tritium permeation through a cooling plate considering a set of different parameters that affect the breeding blanket (BB) performance. Different BB configurations are studied varying the geometrical parameters defining the plate: the wall thickness and the cooling channel pitch. Moreover, the analyses cover every permeation regime, from diffusion-limited to surface-limited, including intermediate situations. This way, results are applicable to every BB concept which employs cooling plates with independence of their specific materials, temperatures and tritium concentrations. Results have been compared with those of 1D models in order to define form factors that can be used to increase the precision of system level tritium transport models. A total of 91 simulations have been carried out to find correlations that allows computing the form factors in a wide variety of situations. Results show that the accuracy of 1D models and consequently the accuracy of most system level models strongly depends on the permeation regime and can be poor in some conditions. However, this study demonstrates that the use of form factors can be used for an effective increase of system level exactness.

On the stability and stationarity of the Super H-mode combined with an ion transport barrier in the core

The Super H-mode, a regime with high pedestal pressure and stored energy, is explored on DIII-D and combined with an ion transport barrier in the plasma core to increase performance. A significant improvement of ion temperatures and confinement is facilitated by favorable conditions such as high rotational shear and high ion pedestal temperatures. As a result of a rise in density and simultaneous decrease in rotation, the ion transport barrier disappears during the discharge evolution, leading to a transition from a very high confinement state at early times, to a reduced but still high confinement phase. Additionally, in many discharges, a global magnetohydrodynamic (MHD) event consistent with the coupling of a destabilized internal mode to an edge localized mode causes a large energy loss and leads to a reorganization of the plasma into a lower temperature, higher density state. Depending on the magnitude of the global MHD event, the plasma edge collisionality can increase significantly and shift the operational boundary from the peeling to the ballooning side, which can be understood as a drop out of the Super H-mode channel into standard H-mode. Hence, in Super H-mode discharges with ion transport barriers, both the improved pedestal height and rotational shear contribute to the high stored energy. At very low levels of rotation, the confinement factor for SH modes is still expected to exceed standard H-mode by 20%–30%. With their overall stationarity and high-performance levels, Super H-mode discharges provide an attractive regime for ITER and may enable a more compact design of future fusion power plants.

Impacts of sawtooth crashes on tokamak plasmas in DEMOs

Self-consistent simulations of sawtooth oscillations on four DEMO (demonstration fusion power plant) designs, proposed by European, Indian, Chinese, and Korean teams, are carried out using the Porcelli sawtooth triggering model and the modified version of the Kadomtsev magnetic reconnection model coupled with the BALDUR integrated code. The simulation results suggest that all sawtooth crashes are triggered by the driving force for the internal kink instability overcoming the fast ion stabilization found in all DEMOs. Conversely, different expansions of the pre-crash helical flux are found, causing different mixing radii on each design. The mixing radius of the European DEMO is found to be the largest, resulting in the largest reduction of the core plasma temperature after the sawtooth crash. It is also observed that for all DEMOs except the Chinese DEMO, regular sawtooth crashes lead to an increase in the core plasma density and radiative power density in different levels, which are due to the steepness of the pre-crash hollow density profile. Furthermore, while helium is found to be outwardly transported by sawtooth crashes for all DEMOs, the sawteeth yield a penetration of the seeded impurities, remarkably found in the European design. Interestingly, based on the European design, a decrease in the atomic mass of the plasma fuels and seeded impurities are found to yield greater flatness of the pre-crash density profile and thus a smaller impact of sawtooth oscillations on central density. Additionally, the sawteeth degrade the plasma performance in the Chinese DEMO, but demonstrate positive impacts on the performance in the European, Indian, and Korean DEMOs, which results from the increase of the central ion density caused by the sawteeth. Furthermore, variation in neutral beam injection power exhibits both positive and negative correlations of sawtooth oscillation impacts on central ion density. The sensitivity of the mixing region of DEMOs is also investigated for scans where the main plasma parameters are varied.

Design Issues for Fusion Commercialization

The EUROfusion roadmap for fusion research was recently updated, and it describes a clear set of missions and associated goals on the route to commercial fusion electricity. Beyond ITER, the main target of the program is the development of DEMO, a fusion technology demonstrator which will produce a substantial net electrical output, breed its own fuel, and demonstrate supporting technologies such as automated remote handling systems aimed for high availability. Work on DEMO has already proven extremely valuable in identifying substantial design integration issues and system interdependencies, which uniquely complicate the fusion power plant design. However, the uncertainties which arise from the low-technology readiness levels of fusion systems mean that DEMO must be robustly designed with substantial margins in the performance, and while it will demonstrate the technological feasibility of an integrated fusion power plant, further work will be required to refine the concept toward attractive commercialization. Under EUROfusion Mission 7, work is now turning toward the wider problems of how fusion-produced energy can be turned into economically viable electrical energy. A fusion power plant is a uniquely challenging environment and requires specialized technologies and materials. It will be important to find crossover applications outside fusion and other ways to ensure reduced costs as they are scaled to full commercial roll-out. This article outlines the EUROfusion approach to solving these problems. It describes the problems faced in engineering a fusion power plant; supply chain and procurement issues that have to be solved; and suggests ways in which fusion power can be made commercially attractive.

Numerical investigation of Al<SUB align=right>2O<SUB align=right>3 nanoparticle effect on a boiling forced swirl flow field: application to the ITER project

One of the most important issues in design of the nuclear fusion power plants is the heat removal from the hottest region at the divertor. Various methods could be employed in order to improve the heat transfer efficiency such as generating turbulent flow and injection of nanoparticles in the host fluid. In the current study, water/Al2O3 nanofluid forced swirl flow boiling has been investigated by using homogeneous thermophysical model within Eulerian-Eulerian framework through a twisted tape tube and boiling phenomenon was modelled using the Rensselear Polytechnic Institute approach. In addition to comparing the results with the experimental data and their reasonable agreement, it was evidenced that higher flow mixing results in more uniform bulk temperature and lower wall temperature along the twisted tape tube. The presence of Al2O3 nanoparticles in the boiling flow field showed that increasing the nanoparticle concentration leads to a reduced vapour volume fraction and wall temperature.

Sensitivity analysis of fusion power plant designs using the SYCOMORE system code

The next step after ITER is the demonstration of stable electricity production with a fusion reactor. Key design performance will have to be met by the corresponding power plant demonstrator (DEMO), fulfilling a large number of constraints. System codes such as SYCOMORE, by simulating all the fusion power plant sub-systems, address these questions. To be able to perform design optimizations, simplified models relying on physical and technological assumptions have to be used, resulting in a large number of input parameters. As these parameters are not always exactly known, the impact of their associated uncertainties on final design performance has to be evaluated. Sensitivity methods, by measuring the relative influence of inputs on the figures of merit of the design, allow the dominant parameters to be selected. This information helps the search for optimal working points, guides the priority for technical improvements and finally allows meaningful inputs for uncertainty propagation to be selected. A full set of sensitivity methods and their application to an ITER and DEMO design will be presented, discussing both the statistical method behaviours and the physical results. Plasma shape parameters (minor radius and plasma elongations) share half of the net electricity power sensitivity for the DEMO 2015 design, while the toroidal magnetic field and the 95% safety factor are responsible for 23% and 17% of the electric power sensitivity, respectively. The plasma minor radius is responsible for 45% of the pulse duration sensitivity for the DEMO 2015 design, while plasma physics parameters drive % of the pulse duration sensitivity.

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.

Activation foil measurements at JET in preparation for D-T plasma operation

Neutronics experiments are carried out at the Joint European Torus (JET) in preparation for the forthcoming high performance Deuterium-Tritium (D-T) campaign. Among others, activation foil measurements are performed in order to evaluate neutron streaming in various positions within the torus hall far from the plasma source. In the present study, the neutron fluence measurements performed using activation foils during the JET Deuterium-Deuterium (D-D) campaign are discussed. The activation foil results in terms of reaction rates are compared against Monte Carlo simulations and a satisfactory agreement is observed. Moreover, the pre-analysis of the experiments to be performed in the forthcoming JET Tritium-Tritium (T-T) and Deuterium-Tritium (D-T) campaigns is presented. The results of the study provide important data and support the preparation of the experimental activities in view of the planned JET D-T operation, allowing the exploitation of the unique 14 MeV neutron yields anticipated. Furthermore, they contribute to the validation of the tools employed in nuclear analyses, which are fundamental for the design and safety of ITER and future fusion power plants.