Fusion Power Reactor Research Articles

The fuel supply quandary of fusion power reactors

ABSTRACT Concern about adequate tritium resources to start up deuterium-tritium (DT) fusion reactors in the near term is misplaced: Proposed demonstration reactors and pilot plants will not realistically be buildable for a great many decades. More disturbing, tritium fuel self-sufficiency of eventual DT reactors may be unattainable, because it is unlikely that tritium-breeding blankets can completely replenish burned-up fuel as well as unburned tritium that is lost in numerous reactor subsystems. The concept of doubling time is unworkable. The only viable fuel solution for sustained operation demands deuterium-based reactors where the tritium is concurrently generated and burned up in the fusioning plasma.

Conceptual design study of the tritium breeding unit for verifying the long-term performance of the breeding blanket

Fusion reactors produce high-energy neutrons through the Deuterium-Tritium (DT) reaction. The long-term performance verification and structural integrity assessment of the breeding blanket under continuous plasma operation are essential for the development of Demonstration fusion power reactor (DEMO) breeding blankets. The Korea Institute of Fusion Energy (KFE) has been planning the construction of a test facility called the Integrated Breeding Test Facility (IBTF) that uses a linear accelerator as the neutron source. A solid beryllium target has been considered to generate fusion-like neutrons. A conceptual design of the Tritium Breeding Unit (TBU), which is a key component of the breeding blanket, has been carried out. This study evaluated tritium production under fusion-like conditions using factors such as the position and thickness of the graphite block, the thickness of the first wall, and the radial length of the breeding zone. The results showed that increasing graphite block thickness and optimizing the dimensions of the TBU enhanced tritium production. However, these changes affected the structural integrity of the TBU. The side wall of the TBU experienced membrane plus bending stress, which exceeded the Level C and D acceptance criteria under the assumption of In-Box LOCA. Measures were investigated to ensure structural integrity while improving tritium production of the TBU.

Fast generation of parametric neutronic models for stellarators. Coupling HeliasGeom and GEOUNED

The operation of the Wendelstein 7-X has highlighted stellarators as potential fusion power reactors. As a consequence, the pre-conceptual design of the next HELIAS stellarator has started in the framework of EUROfusion. During these first stages, it is essential to assess the feasibility of various stellarator concepts. It is important to emphasize that these concepts can vary significantly in terms of their overall shape and allocation of space for each component, including the vacuum vessel, breeding blankets, among others.Although pre-conceptual neutronic analysis can be conducted using straightforward yet representative neutronic models, such as simplified multilayer models with minimal geometric details, the intricate shape of stellarators presents challenges to produce them quickly. This issue prevents the effective parametric optimization of the concepts, creating a bottleneck in the design of HELIAS.In order to overcome this problem, we are developing a computational workflow for the fast production of parametric stellarator neutronic MCNP models. This work presents the current status of the scheme composed of two steps: HeliasGeom for the production of parametric multilayer CAD models, and GEOUNED for the translation of these models to MCNP CSG neutronic models.

Study on hot isostatic pressing conditions of ARAA for fabrication of the breeding blanket first wall

The helium-cooled concept with solid breeding materials is one of the candidates for demonstration fusion power reactor (DEMO) breeding blanket. The first wall (FW) is a core component of the breeding blanket. The FW should keep the structural integrity during the normal plasma operation which is subject to high surface heat flux, neutron wall load, and high coolant pressure simultaneously. Various fabrication methods have been studied for the development of the FW fabrication technologies. This study focuses on FW fabrication technologies with hot isostatic pressing (HIP) bonding which is considered as the most promising method for achieving such complex shape like FW having internal cooling channels. We investigated the HIP bonding conditions using Advance Reduced Activation Alloy, ARAA, which is a reduced activation ferritic-martensitic steel under development in Korea. Temperature, holding time and pressure were considered as the parameters to optimize the HIP bonding conditions. The microstructures around the HIP joints changed with respect to the HIP bonding temperature. The amount of oxide particles around the HIP joints varied depending on the degassing conditions. The tensile strength of the HIP joint was not greatly affected by the presence of the HIP bonding line and degassing conditions. On the other hand, it has been confirmed that the Charpy absorbed energy of the HIP joint was significantly affected by degassing conditions. An increase in oxide particles around the HIP joints occurred when degassing temperature and vacuum level were low, and as the quantity of oxide particles increased, the Charpy absorbed energy of the HIP joint decreased substantially.

An optimization of efficient combined cycle power generation system for fusion power reactor

Fusion power plants can meet the energy demands of the world. The high thermal performance of a power generation system is still a challenge and one of the developing trends with a fusion reactor due to the high outlet temperature. A combined cycle system has been optimized to work at high outlet temperature and pressure of fusion power reactor with thermal power of 2500 MWth and its input parameter has been optimized along with impact of partial load to achieve high thermal performance very first time. The combined cycle based on the concept of two stages of expansion in the closed-Brayton-cycle and two reheaters in the Rankine cycle named Schematic-IV with the higher thermal performance of 58.19% as compared to Schematic-I, II and III has been proposed for a fusion power reactor. The increase in more than one expansion stage and reheat stage is not economical because it increased the thermal performance by about half of performance as compared to one. The effect of isentropic efficiency of compressor and heat rate have been investigated to validate the thermodynamic model and calculations. The optimized thermal performance of the combined cycle is 58.19% at a feed water mass flow rate of 592 kgs−1 and steam outlet temperature of 277 °C in the combined cycle. The heat rate decreased and thermal performance increased with the mass flow rate verifying that the thermodynamic model is correct. The thermodynamic analysis of the combined cycle-based nuclear reactor provides insights for better system thermal performance and reveals the effect of key parameters on the system performance.

Development of radiation sources based on CAD models for the nuclear analysis of IFMIF-DONES lithium loop

IFMIF-DONES is a neutron source designed to irradiate materials to be used in future fusion power reactors such as DEMO. The facility is based on a deuteron beam impinging onto a liquid lithium jet to generate the neutron flux. Lithium and Corrosion Products will get activated and produce Be-7 and Activated Corrosion Products (ACP). These products will be distributed along the lithium loop, both dissolved in lithium and deposited locally. This complex gamma source should be properly represented to perform radiological safety studies. A new tool, called CAD2CDGS, was created to represent these sources. This tool creates decay gamma sources based on CAD geometries and codifies them into CDGS format. Sources are specified in the CAD model, allowing a user-friendly approach. The tool is based on Open-Source tools (FreeCAD) and EUROfusion codes (cR2S European SDDR tool). In this article the CAD2CDGS tool methodology, inputs and workflow are explained. Two verification tests have been done to check the correctness. Finally, it is demonstrated the applicability of the tool to IFMIF-DONES, with the radiological zoning analysis of the rooms surrounding the Lithium Loop Cell impact by the Be-7 and ACP.

DONES EVO: Risk mitigation for the IFMIF-DONES facility

The International Fusion Materials Irradiation Facility- DEMO Oriented Neutron Source (IFMIF-DONES) is a scientific infrastructure aimed to provide an intense neutron source for the qualification of materials to be used in future fusion power reactors. Its implementation is critical for the construction of the fusion DEMOnstration Power Plant (DEMO).IFMIF-DONES is a unique facility requiring a broad set of technologies. Although most of the necessary technologies have already been validated, there are still some aspects that introduce risks in the evolution of the project.In order to mitigate these risks, a consortium of companies, with the support of research centres and the funding of the CDTI (Centre for the Development of Industrial Technology and Innovation), has launched the DONES EVO Programme, which comprises six lines of research:•Improvement of signal transmission and integrity (planning and integration risks)•Optimisation of RF conditioning processes (planning and reliability risks)•Development of a reliable beam extraction device (reliability risks)•Development of technologies for the production of medical isotopes (reliability risks)•Improvement of critical parts of the lithium purification system (safety and reliability risks)•Validation of the manufacture of critical components with special materials (reliability risk).DONES EVO will focus on developing the appropriate response to the risks identified in the IFMIF-DONES project through research and prototyping around the associated technologies.This contribution will present a discussion of the risks, the proposed response to them and the evolution of the technologies involved, following selected experiments carried out throughout the project.

Relationship between magnetic field and tokamak size—a system engineering perspective and implications to fusion development

High temperature superconductors (HTSs) offer the promise of operating at higher magnetic field and temperature. Recently, the use of high field magnets (by adopting HTS) has been promoted by several groups around the world, including new start-up entries, both to substantially reduce the size of a fusion power reactor system and as a breakthrough innovation that could dramatically accelerate fusion power deployment. This paper describes the results of an assessment to understand the impact of using high field magnets in the design of DEMO in Europe, considering a comprehensive list of physics and engineering limitations together with the interdependencies with other important parameters. Based on the results, it is concluded that increasing the magnetic field does not lead to a reduction in device size with relevant nuclear performance requirements, because (i) large structures are needed to withstand the enormous electromagnetic forces, (ii) thick blanket and n-shield structures are needed to protect the coils from radiation damage effects, and (iii) new divertor solutions with performances well beyond today’s concepts are needed. Stronger structural materials allow for more compact tokamaks, but do not change the conclusion that scalability is not favourable when increasing the magnetic field, beyond a certain point, the machine size cannot be further reduced. More advanced structural support concepts for high-field coils have been explored and concluded that these solutions are either unfeasible or provide only marginal size reduction, by far not sufficient to account for the potential of operating at very high field provided by HTS. Additionally, the cost of high field coils is significant at today’s price levels and shows to scale roughly with the square of the field. Nevertheless, it is believed that even when not operated at high field and starting within conventional insulated coils, HTS can still offer certain benefits. These include the simplification of the magnet cooling scheme thanks to increased temperature margin (indirect conduction cooling). This in turn can greatly simplify coil construction and minimize high-voltage risks at the terminals.

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.

Tritium Content in Komatsuna Cultivated in Tritium-Contaminated Peat Soil

Tritiated water from fusion power reactors will be the next major issue when fusion technology comes fully onstream. Effective radiation protection measures will be implemented when the scope of its behavior is well understood. To understand tritium behavior in the environment, komatsuna was cultivated in tritium-contaminated peat soil. It was indicated experimentally from water immersion experiments that the amount of tissue free water tritium in komatsuna depends on the tritium concentration in the soil and that the concentration in stems and leaves in komatsuna decreases as the tritium concentration in the soil decreases. The amounts of tritium retained in the roots were much less than that in the stems and leaves.

Molten salts evaluation for application in the fusion power plant's intermediate heat transfer system

One of the challenges of fusion power plants' integration into the power grid is the pulse operation of the first fusion power reactors, which poses critical requirements for construction materials, power conversion systems, and energy storage systems. There are different ways and approaches which aim to solve this pulse operation issue. Among the many proposed solutions is an intermediate heat transfer system with molten salts acting as energy storage and pulse balancing technology.This article aims to evaluate different molten salts, their advantages, disadvantages, the possibility for application, and costs for the intermediate heat transfer system with two tanks design. The properties such as boiling and freezing temperature, heat capacity, density, and more are compared for various salts. The cost of the energy storage systems with chosen salts and two tanks is estimated.

Development of pebble-based extruded carbon rods for extreme plasma heat flux environments

This work presents first experiments toward the development of continuously renewable (extrudable) pebble-based carbon rods for use as plasma-facing components in extreme steady-state plasma flux environments. The primary envisioned application of this work is a first wall that can survive long-term in future magnetic fusion power reactors while also improving recovery of the reactor fuel (tritium and deuterium atoms). Bench tests applying extreme steady-state front-surface heat loads of up to 50 MW/m2 are presented. Continuous pebble rod front-surface recession and intact pebble recovery are successfully demonstrated, at a rate of order 0.2 cm/s. Numerical simulations of the pebble rod front-surface recession are able to match observations reasonably well, indicating that the recession mechanism can be understood as occurring due to pebble thermal expansion and resulting shock and cracking of the inter-pebble binder. Tests of the pebble rod extrusion demonstrate that friction between the rods and the stainless steel extrusion channel is tolerably low (<50 N for the expected channel length) over a wide range of temperatures. Front-surface outgassing rates below 1000 Torr L/s/m2 are achieved, believed to be sufficiently low for use in magnetic fusion reactors. Initial parametric scans over pebble rod size and binder fraction to vary front-surface recession rates are presented.

Concept of Contamination Control Door for DEMO and Proof of Principle Design

During the maintenance period of a future fusion reactor power plant, called DEMOnstration Power Plant (DEMO), remotely handled casks are required to confine and handle DEMO in-vessel components during their transportation between the reactor and the active maintenance facility. In order to limit the dispersion of activated dust, a Contamination Control Door (CCD) is designed to be placed at an interface between separable containments (e.g., vacuum vessels and casks) to inhibit the release of contamination at the interface between them. The remotely operated CCD—technically, a double lidded door system—consists of two separable doors (the cask door and port door) and three different locking mechanisms: (i) between the cask door and cask, (ii) between the cask door and port door and (iii) between the port door and port. The locking mechanisms are selected and assessed according to different criteria, and the structure of the CCD is optimized using an Abaqus Topology Optimization Module. Due to the elastic properties of the CCD, deflections will occur during the lifting procedure, which may lead to malfunctions of the CCD. A test rig is developed to investigate the performance of high-risk components in the CCD in the case of deflections and also malpositioning. Misalignment can be induced along three axes and three angles intentionally to test the single components and items. The aim is to identify a possible range of operating in the case of misalignments. It is expected that the proposed CCD design should be able to operate appropriately in the case of ±3 mm translational misalignments and ±1° rotational misalignments.

Integrated Design of the Vacuum and Safety Barrier between the Lithium and Test Systems of IFMIF-DONES

The international fusion materials irradiation facility-DEMO-oriented neutron source (IFMIF-DONES) is a facility that 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 energy spectrum like the one present in a fusion power reactor. The material specimens are irradiated in a containment structure named the test cell (TC), which is part of the test systems (TS). The TC also houses a part of the other major system (lithium system, LS), which provides the liquid lithium for the reaction through a piping system. At a point, the lithium piping needs to exit the TS, but the primary safety boundary must be continuous around these penetrations. Therefore, a special barrier, called the test systems–lithium systems interface cell (TLIC), has been developed around the piping system to provide a safety-approved and remotely maintainable vacuum boundary envelope. In this paper, the integrated design development of the TLIC is described, consisting of the design development according to the RCC-MRx code, the remote-handling (RH) needs, and the procedures and safety-related special needs of the design.

Rationale Behind EU-DEMO Limiter’s Plasma-Facing Component Design Under Material Phase Change

The protection strategy adopted for the European DEMOnstration (EU-DEMO) fusion power reactor foresees the use of sacrificial components–referred to as limiters–dealing with plasma-wall contacts. Their aim is to protect the first wall (FW) against the huge amount of plasma energy (up to GigaJoules) released in a few milliseconds during disruptive events, which might lead to melting and/or vaporization of the foreseen plasma-facing tungsten armor. The current limiter design concepts rely on actively water-cooled plasma-facing components (PFCs) made of tungsten. As water is not allowed inside the main chamber, limiter’s PFCs must be designed to preserve the cooling system integrity under any scenario, therefore the estimate of the thickness of material undergoing any phase change is crucial. Given the initial assessment of plasma magnetic configurations during off-normal events, this article describes the procedure followed for designing the limiter’s PFC, which includes a novel approach for estimating the molten thickness of material under high heat flux. A simplified 1-D model has been implemented in MATLAB, which deals with multiphase moving boundary problems, hence its name Thermal Analysis foR Tracking InterFaces under meLting&vaporizaTion-induced plasma Transient Events (TARTIFL&TTE). This model takes its inspiration from the way phase change interface tracking problems are tackled in food industry (i.e., freeze-drying processes) and solute concentration diffusion-controlled problems. It overcomes the complexity of solving a strongly coupled non-linear system of partial and ordinary differential equations in moving spatial domains by adopting a change of coordinate system based on the Landau transformation. As a result, an equivalent and fixed spatial coordinate system is defined where the spatial domain boundaries of the different phases are constrained, and the systems of governing equations are easy to solve. Both the solid and liquid phases are modeled, while the vapor is assumed to be removed once formed. Its benchmark against computational results found in the literature has shown a very good agreement, which paves the way to further development of it. For phase change interface tracking problems in 2-D/3-D and more complex geometries, a commercial Multiphysics software adopting the Lagrangian–Eulerian approach in moving mesh frames will be used for tackling problems where material is removed following a phase change. Although the vapor domain is not simulated, a set of gas kinetics boundary conditions couples the interface between vapor and liquid phases, driving its position over time. This will be detailed described in a future companion article.

Operational Tritium Inventories in the EU-DEMO Fuel Cycle

The European Demonstration Fusion Power Reactor (EU-DEMO) has to operate in a completely tritium self-sufficient mode after initial start-up, which includes producing excess tritium to allow the start-up of other reactors. The initial start-up inventory is mainly dictated by operational inventories in the fuel cycle (FC). Advances in FC technologies and immediate recycling of a large fraction of the torus exhaust gas in the direct internal recycling loop are expected to contribute greatly to an overall low operational inventory. The remainder of the torus exhaust gas, as well as tritium from the blankets, nevertheless requires treatment in the tritium plant in order to perform the necessary purification and isotope rebalancing. Here, the employed systems still feature significant operational inventories and predominantly require steady-state operation in order to maximize their performance. In this paper the operational tritium inventories in the major FC systems are reported based on the pre-concept FC design. Additionally, major dependencies of these inventories on key design drivers of the FC are discussed. It is predicted that the EU-DEMO FC will be able to operate with an overall tritium inventory of less than 2 kg.

Nuclear data libraries for IFMIF-DONES neutronic calculations

International Fusion Materials Irradiation Facility-DEMO Oriented NEutron Source (IFMIF-DONES) is an installation aimed to irradiate with a high neutron flux materials relevant for the construction of the DEMOnstration fusion power plant (DEMO), in order to study the damage due to irradiation. Neutrons are generated using a 40 MeV and 125 mA deuteron beam impinging on a thick liquid lithium target. With these characteristics, damage due to irradiation comparable to that in the first wall of a fusion power reactor is achieved. In this paper we investigate the differences in the neutronic calculations of the IFMIF-DONES design when using different nuclear data libraries. We first studied the differences in neutron production due to Li(d, xn) reactions between different models and evaluations, comparing the different results with experimental data. Additionally, we tested the performance of the MCNP6.2 and Geant4 Monte Carlo codes when using deuteron incident data libraries. Then, we performed neutronic calculations of the IFMIF-DONES design using the most reliable Li(d, xn) neutron production models available, which are the FZK-2005 and JENDL/DEU-2020 evaluations according to the results obtained in the first part of the study. Thus, the differences in these evaluations are propagated to different neutronic calculation results: neutron flux, primary displacement damage, gas production, and heating in the materials to be irradiated. Finally, we also carried out these same neutronic calculations while using different nuclear data libraries for the neutron transport.

Development of fabrication technologies for the first wall of the HCCR blanket

The helium-cooled ceramic reflector (HCCR) concept is a primary candidate among demonstration fusion power reactor (DEMO) blanket options in Korea. The HCCR blanket is composed of a first wall (FW), a breeding zone, side cover plates and a back manifold. This study focuses on fabrication technologies for the FW using laser-beam welding (LBW) and hot isostatic pressure (HIP) bonding. Permanent backing strips in the pressure boundaries of the HCCR blanket are not allowed according to regulations related to pressure vessels and nuclear pressure vessels because high-pressure helium gas at 8 MPa is used as the coolant. LBW conditions and procedures for the fabrication of a FW without backing strips were investigated using small mock-ups with an advanced reduced-activation alloy called ARAA, which was developed in Korea. The laser beam power, thickness of the cover plate, and width of the backing strips were considered as parameters. The LBW conditions and procedures were optimized to remove the backing strip and achieve uniform and continuous back beads. After LBW, the fabrication of the FW with small mock-ups was completed by joining the back plate with HIP bonding. The small mock-ups were subject to very high temperatures exceeding 1000°C during the LBW and HIP bonding processes, which led to changes in the microstructures and hardness levels. The microstructures and hardness levels were recovered through post-heat treatments performed after HIP bonding. The microstructures of the HIP bonding interface after the heat treatments were observed. A partial HIP bonding line was confirmed and transverse tensile tests were carried out to investigate the impact of the partial HIP bonding line on the mechanical properties of the HIP bonding interface. The tensile strength of the HIP bonding interface was approx. 2% lower than that of the base metal, indicating that the partial HIP bonding line caused a slight decrease in the tensile strength, but the impact on the HCCR blanket design would be negligible.

Engineering design of the CFETR machine

The China Fusion Engineering Test Reactor (CFETR) is the next step in China's roadmap for magnetic confinement fusion development. It aims to bridge the gap between the ITER fusion experimental reactor and the demonstration fusion power reactor (DEMO). The CFETR will be operated in two phases. Phase I aims to achieve fusion power of 50–200 MW, Qplasma = 1–5, TBR > 1.0 and neutron irradiation effects ∼ 10 dpa. Phase II aims to perform tokamak DEMO validation under fusion power > 1 GW, neutron irradiation effects ∼ 50 dpa and figQplasma > 10. To ensure that the CFETR machine can achieve these goals, the major/minor radii of plasma have been altered from R = 5.7 m/a = 1.6 m to R = 7.2 m/a = 2.2 m. The new machine can meet the targets of both Phase I and Phase II and achieve smooth transfer between them. The configuration of the CFETR machine is presented in this paper. Then, the design progress and status of some key components are described, including the magnet system, blanket system, diverter system, remote handling system, vacuum vessel, thermal shield and cryostat. The research and development of most components are underway and will be demonstrated in the near future.

MIRA: a multi-physics approach to designing a fusion power plant

Fusion systems codes (SCs) are deployed to produce the baseline of the European fusion power reactor (DEMO) within its conceptual design. A DEMO baseline is mostly defined by a radial/vertical reactor sketch and major reactor parameters, such as fusion and net electric power, magnetic fields, and plasma burn time. A baseline shall also meet a set of prescribed reactor requirements, constraints, and architectural features. According to the conceptual design workflow implemented within the EU-DEMO programme, the output from the SC is transferred to the detailed physics and engineering design codes. Presently-available fusion SCs rely on rather basic physics and engineering models (mostly at zero or one-dimensional level). The design codes, instead, are very detailed but run on much longer computing times. To fill the gap between systems and design codes, the multi-fidelity systems/design tool modular integrated reactor analysis (MIRA)—has been recently developed. MIRA incorporates the physics and the engineering insights of the utmost domains of tokamak reactors and relies on a higher spatial resolution, spanning from 1D up to 3D modelling frames. The MIRA approach has been applied to the DEMO 2017 baseline, generated by the EU reference SC PROCESS and used as input to MIRA. In the paper, the architectural and mathematical insights of the MIRA package are described, along with an EU-DEMO 2017 baseline analysis.