DEMO Fusion Reactor Research Articles

Production and transport modelling of Po-210 in DEMO reactor

One of the generic designs of the nuclear fusion DEMO reactor proposed by the EUROfusion consortium foresees the development of a tritium breeding blanket (BB) relying on the use of the liquid-metal PbLi eutectic alloy as both neutron multiplier and tritium breeder, namely the water-cooled lithium lead (WCLL) BB, whose strengths and weaknesses are well known. This paper focuses the attention on one of the possible disadvantages of such a technology: the production of the highly radiotoxic radionuclide 210Po, which could become a safety issue to be accounted for. The 210Po concentration within the PbLi circuit has been assessed by solving a modified version of Bateman’s equations to consider the alloy circulation, so a one-dimensional convective fluid-dynamic model has been set up. Nuclear quantities have been evaluated by Monte Carlo neutron transport analyses using MCNP code and adopting a fully heterogeneous model of DEMO equipped with the WCLL BB. Moreover, rough sensitivity analyses have been performed to assess the influence on the results of the uncertainties related to the 209Bi radiative-capture cross section and the initial concentration of this nuclide which is present in the PbLi as an impurity. Results obtained have been critically discussed and some safety issues have been addressed to evaluate the possible hazard in case of a leak of PbLi accident.

Identification of the Postulated Initiating Events of Accidents of a CPS-Based Liquid Metal Divertor for the EU DEMO Fusion Reactor

A preliminary but systematic safety analysis of a liquid metal divertor (LMD) for the EU DEMO performed by means of the Functional Failure Mode and Effect Analysis (FFMEA) is presented. This methodology is suitable for the analysis of the LMD, which is undergoing preconceptual design. In fact, the FFMEA compensates for the lack of detailed design information by postulating the loss of a system function, rather than a specific component failure. The implementation of the FFMEA led to a better understanding of the safety and operational issues associated with the system and to the identification of a list of postulated initiating events (PIEs), i.e., the most challenging conditions for the plant. The PIEs, together with their possible consequences, represent an input for future quantitative safety analyses. Due to the early design stage of the LMD and the iterative nature of the methodology, this list will evolve alongside the design detail and with improvements in the understanding of phenomena driving reactor behavior. The study highlighted some safety-relevant issues, e.g., those related to materials compatibility and system modularity, to be addressed in the perspective of a safety-driven design evolution.

A planning study for virtual DEMO development in Korea

We report a planning study to develop an integrated framework, on which a fusion power plant can be virtualized and simulated. As an effort to develop software for designing Korean DEMO (K-DEMO) fusion reactor, key technologies are identified with a structure to organize and integrate them. Strategies and directions for the development are proposed considering the Korean national plan to invest on the 4th industrial revolution technologies such as digital twin, artificial intelligence etc. Also, a road map for the development is proposed, in which 4 development phases are established in close connection to the Korean fusion energy R&D plan. Starting with the development of Virtual KSTAR (V-KSTAR) in the first phase, the roadmap proposes to develop Virtual DEMO (V-DEMO) in the final phase. The deliverables from each development phase are defined with discussions on their applications to the K-DEMO design activities.

European Demo Fusion Reactor: Design and Integration of the Breeding Blanket Feeding Pipes

European Demo Fusion Reactor: Design and Integration of the Breeding Blanket Feeding Pipes

Power exhaust concepts and divertor designs for Japanese and European DEMO fusion reactors

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan–Europe broader approach DEMO design activity (BA DDA). Common critical issues are the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling (P rad tot/P heat ⩾ 0.8). Different exhaust concepts in the main plasma and divertor have been developed for Japanese (JA) and European (EU) DEMOs. JA proposed a conventional closed divertor geometry to challenge large P sep/R p handling of 30–35 MW m−1 in order to maintain the radiation fraction in the main plasma at the ITER-level (f rad main = P rad main/P heat ∼ 0.4) and higher plasma performance. EU challenged both increasing f rad main to ∼0.65 and handling the ITER-level P sep/R p in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding P sep = 250–300 MW and 150–200 MW, respectively. Both results showed that large divertor radiation fraction (P rad div/P sep ⩾ 0.8) was required to reduce both peak q target (⩽10 MW m−2) and T e,i div. In addition, the JA divertor performance with EU-reference P sep of 150 MW showed benefit of the closed geometry to reduce the peak q target and T e,i div near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130 °C–200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs.

R&D Activities for Fusion DEMO in the QST Rokkasho Fusion Institute

The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokkasho Fusion Institute is reported. In order to advance the Japanese DEMO activity, not only Japanese domestic activity but also international collaborations of Broader Approach activity and ITER-related activities are conducted in the QST Rokkasho Fusion Institute. Activities for DEMO design and relevant R&D; design of a fusion neutron source and development of an accelerator, ITER Test Blanket System; tritium handling technology; and information technology infrastructures, including a supercomputer system and a remote experimentation system, are carried out for a Fusion DEMO reactor.

Optimization of the overall Toroidal Field Coil cryomagnetic system at the pre-conceptual design phase of the European DEMO fusion reactor

Different superconducting conductor design options are considered for the Toroidal Field (TF) Coil magnets of the future European DEMO reactor. Four TF Winding Pack (WP) options have been proposed and consist of two layer-wound concepts and two pancake-wound concepts, each of them having specific conductor design. These TF WPs are cooled by supercritical helium forced flow driven by a cold circulator at about 4 K. At the pre-conceptual design stage, it is particularly relevant to also consider cryogenic system merits while designing the TF magnet system. Indeed, both cost drivers such as the superconductor amount and the refrigeration loads depend on the cooling parameters: temperature, pressure, mass flow/pressure drop across the conductor. Hence, these parameters shall be optimized in order to minimize the total cost for the cryomagnetic system. For the pancake-wound option designed by CEA, a new approach that combines the magnet design calculation and the refrigeration load calculation using CEA tools (an adapted version of MADMACS-TF code and the simulation library Simcryogenics) was implemented. An optimization problem that considers both magnet and cryogenic merits was assessed. This problem consists in optimizing the superconducting amount and the refrigeration loads while ensuring that the temperature margin of the superconductor remains at 1.5 K. This article presents the method for cost minimization and the information workflow between the different codes, as well as preliminary results. This method is relevant to optimize simultaneously the design of the magnets and the cryogenic system and to make comparisons between the different conductor designs during the next phase of EU DEMO, moving forwards to the conceptual design. With such a method, recommendations can be made to ensure that the combined investment and energetic consumption cost is kept as low as possible.

Large Area Diamond Disk Growth Experiments and Thermomechanical Investigations for the Broadband Brewster Window in DEMO

An electron cyclotron system (ECS) is required in the European DEMO fusion reactor for plasma heating, magnetohydrodynamic control, especially the suppression of neoclassical tearing modes (NTMs), and thermal instability control. Microwave beams of 2 MW, generated by gyrotrons, shall propagate through corrugated HE <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">11</sub> waveguides (WGs) of 63.5-mm inner diameter and finally reach specific locations in the plasma by a dedicated set of mirrors. The deployment of the beam may be done either by angular steering, i.e., tilting moveable mirrors, or by frequency steering, i.e., tuning the gyrotron frequency. In the last option, for long pulse operations, broadband optical grade diamond windows are required as confinement and/or vacuum barriers both at the torus and the gyrotron side. The Brewster-angle window represents a promising solution due to its intrinsic simplicity. However, the WG aperture imposes a minimum Ø180 mm for the disk at the Brewster-angle of 67.2° for diamond. This is an ambitious target, considering 140 mm as the state of the art. First, this article shows the very promising results of diamond growth experiments carried out by Diamond Materials GmbH with the first worldwide free-standing optical grade Ø180-mm disk having an average unpolished thickness of 1.3 mm. For the first time, one side of this large disk was polished and loss tangent measurements were performed at the center and over a large area. Then, the performance of the window featuring these large disks was characterized by a consistent set of thermomechanical analyses in view of the forthcoming window prototyping activity.

Thermal-hydraulic analysis of the DTT CS and PF pulsed coil performance during AC operation

The EU DEMO fusion reactor, now entering in its conceptual design phase, will have to tackle several challenges, such as the long plasma pulse duration and the exhaust of very high heat fluxes from the plasma. The Divertor Tokamak Test (DTT) facility, a compact superconducting tokamak under construction at ENEA Frascati, will address the power exhaust problem in DEMO by testing several DEMO-relevant divertor solutions and operation scenarios.The DTT superconducting coils are the subject of the analyses presented here: as the tokamak must be very flexible to face different plasma scenarios, the design of the magnet system must be proven to be robust. In particular, the thermal-hydraulic performance of the coils operated in pulsed mode, namely the Central Solenoid (CS) and the Poloidal Field (PF) coils, will be analyzed. The CS is composed of 6 modules, layer-wound with Nb3Sn Cable-in-Conduit Conductors (CICCs), while the PF coils are pancake-wound. The two PF coils closest to the machine axis are wound using Nb3Sn CICCs, while the others adopt NbTi CICCs. All the coils are cooled by supercritical He (SHe) in forced flow.The pulsed operation of these coils induces AC losses, eroding their temperature margin: a detailed thermal-hydraulic model of the DTT pulsed coils is developed here using the 4C code. The model is then applied to the simulation of the two reference plasma scenarios, a single null and a double null, computing the minimum temperature margin. Different cooling options (single- or double-pancake) are investigated for the PF coils, while the sensitivity to the coupling time constant value is assessed for the CS.

Design status of the Vacuum Vessel of DTT facility

The Divertor Tokamak Test (DTT) facility is an experimental facility under design and construction at ENEA C.R. Frascati. The aim of DTT is to investigate the power exhaust problem in a tokamak, providing possible alternative divertor solutions, with respect the conventional one, which can be extrapolated to DEMO fusion reactor. One of the main components of DTT facility is the Vacuum Vessel (VV), which has the function of providing an enclosed, vacuum environment for the plasma, acting also as a first confinement barrier. Starting from geometrical constraints, imposed by the desired plasma scenario and the configuration needed for the magnetic coils, the conceptual design of the VV was developed and dedicated research activities were carried out to verify design choices. A multiphysical approach was adopted with the aim of identifying a feasible and reliable solution for the VV, taking into account functional and design requirements, relevant aspects and issues. In particular, to assess and to optimize the VV design, fluid-dynamic, thermal and structural analyses were carried out, which are presented in the paper, as well as the progresses of the design.

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.

Statistical Analysis of Tritium Breeding Ratio Deviations in the DEMO Due to Nuclear Data Uncertainties

For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. The uncertainty of the TBR in the neutronic simulations includes many different aspects such as the uncertainty due to the simplification of the geometry models used, the uncertainty of the reactor layout and the uncertainty introduced due to neutronic calculations. The last one can be reduced by applying high fidelity Monte Carlo simulations for TBR estimations. Nevertheless, these calculations have inherent statistical errors controlled by the number of neutron histories, straightforward for a quantity such as that of TBR underlying errors due to nuclear data uncertainties. In fact, every evaluated nuclear data file involved in the MCNP calculations can be replaced with the set of the random data files representing the particular deviation of the nuclear model parameters, each of them being correct and valid for applications. To account for the uncertainty of the nuclear model parameters introduced in the evaluated data file, a total Monte Carlo (TMC) method can be used to analyze the uncertainty of TBR owing to the nuclear data used for calculations. To this end, two 3D fully heterogeneous geometry models of the helium cooled pebble bed (HCPB) and water cooled lithium lead (WCLL) European DEMOs were utilized for the calculations of the TBR. The TMC calculations were performed, making use of the TENDL-2017 nuclear data library random files with high enough statistics providing a well-resolved Gaussian distribution of the TBR value. The assessment was done for the estimation of the TBR uncertainty due to the nuclear data for entire material compositions and for separate materials: structural, breeder and neutron multipliers. The overall TBR uncertainty for the nuclear data was estimated to be 3~4% for the HCPB and WCLL DEMOs, respectively.

Status of the engineering design of the IFMIF-DONES Central Instrumentation and Control Systems

This work presents the current status of the design of Central Instrumentation and Control Systems for the IFMIF-DONES plant. IFMIF-DONES is the International Fusion Materials Irradiation Facility-DEMO Oriented NEutron Source, an accelerator-based neutron source producing a high-energy deuterons beam focused on a fast flowing liquid lithium jet to generate high-energy neutrons via stripping reactions. The intensity and irradiation volume is suitable to generate material irradiation data for design, licensing, construction and safe operation of the DEMO fusion reactor. After a general overview of the current status of the design of the overall architecture, a particular attention is given to the data management including the EPICS control framework, the general rules and criteria for the design and implementation of the HMI system, and the timing system solution composed of a three-layer distribution network to deliver time at each element of the plant that implements an optimal trade-off between timing performance and cost. Eventually, the next design steps and future work are outlined.

The influence of the long-term heating under H2 atmosphere on the tritium release behavior from the neutron-irradiated Li2TiO3

Solid tritium breeding materials are expected to be used in high-temperature conditions for a long time in a fusion DEMO reactor. Thus, it is important to understand how bred tritium releases from the long-term heated material under the environment as close to a DEMO condition as possible to establish a tritium fuel cycle and keep it safe. In this work, the tritium release behavior from the irradiated Li2TiO3 pebbles that was preheated for 720 h at most was observed by heating at 1000 °C under 1000 Pa H2/Ar gas flow. The release peaks of HTO and HT were observed around 300 °C due to the desorption of the chemisorbed water. Also, the broad HTO peak was observed in a higher temperature region despite purging H2/Ar gas. This result suggests that this tritium was released without exchanging with H2 but with combining with oxygen in the pebbles. Moreover, the released tritium amount decreased as the pre-heating time. Finally, the amount of tritium that could not be released by the heating experiment was quantified by dissolving samples with an acid solution. Besides, the total tritium release ratio was discussed.

Performance Optimization of Power Generation System for Helium Gas/Lead Lithium Dual-coolant Blanket System

Helium gas (He-gas) / Lithium Lead (LiPb) Dual-cooled LiPb (DLL) blanket is one of the promising candidates for the DEMO fusion reactor for power generation. Brayton cycle based on He-gas has been used as the power generation system for the fusion and fission power reactors. In this work, a modified and efficient design of the power generation system with concept of two stage expansions and one intercooler named as Schematic-I has been proposed for the DLL blanket of Fusion Power Reactor (FDS-II) to increase the thermal performance (e.g. power density and efficiency) of the system and optimized the parameters of the system. The engineering equation solver tool was used to calculate the performance of the system. It is found that the use of more than one expansion stage with proper selection of intercooler is very effective as compared with either splitting the flow equally for the expansion in two turbines or a single stage expansion. The concept of two-stage expansions in the Schematic-I increased the thermal efficiency of the system up to 51%, which was 4% higher than previous design of DLL blanket power generation system. The main parameters of the system such as isentropic efficiency of turbine and compressor, pressure ratio, recuperator performance, the terminal temperature difference of the recuperator, the inlet temperature of DLL blanket and effect of different pressures on the system have all been analyzed and optimized to achieve high performance.

Development of a compact real-time process gas analysis system for tritium accountancy for a DEMO fusion reactor by an application of laser Raman spectroscopy

In the fuel cycle of ITER and the future DEMO fusion reactors, which has many processes of tritium provision/ consumption/recovery/loss, such as, fueling, DT burning, vacuum pumping, purification, isotope separation and storage, detritiation, etc., tritium accountancy control will be required in its operation and safety management. In contrast to ITER, which will have frequent intervals between operations, a DEMO fusion reactor generating electricity must be operated continuously for long periods. This requires a dynamic tritium accountancy control, especially at the mass balance area boundary of the Tokamak including the vacuum vessel. To make a dynamic tritium accountancy system for a DEMO fusion reactor, based on our past work carried out during the ITER EDA, we are developing a new compact hydrogen isotope gas analysis system by the application of laser Raman spectroscopy advanced by technological progress in laser and detection systems. A prototype system has been made at TPL using a semiconductor laser with a Peltier-cooled CCD detector, with a simple optical system. Results of some preliminary tests being conducted to confirm sensitivity and accuracy under static and dynamic conditions are presented.

Characterization of aluminum-based coatings after short term exposure during irradiation campaign in the LVR-15 fission reactor

Protective aluminum-based coatings represent a promising anti-permeation and anti-corrosion barrier for breeding blanket systems developed for European DEMO fusion reactor. Following the prior in-depth characterizations, selected coating candidates were subjected to a combined test consisting of contact with liquid Pb-16Li, its repeated in-situ solidification and re-melting, tritium permeation during gamma and neutron irradiation in the LVR-15 fission reactor. During the sample exposure in the reactor, the temperature of Pb-16Li was between 300 and 425 °C. The irradiation damage averaged over the sample volume estimated by FISPACT code was limited to 0.037 dpa. This article presents post-irradiation characterization of cylindrical Eurofer97 samples coated by electro-chemical X-metal deposition from ionic liquid (ECX) and pulsed laser deposition (PLD) techniques. The coating damage relative to a reference uncoated sample is discussed.

A design proposal for the European DEMO superconducting bus bars and current leads

The integration of the feeder system in a tokamak involves complex engineering, manufacturing and logistics issues. Therefore, within the pre-conceptual design activities of the European DEMO fusion reactor, it is important to provide a preliminary sizing of the feeders at an early stage. Here we present the conceptual design of superconducting bus bars and current leads for the DEMO feeder system. The proposal is based on three main components: Nb3Sn lead extensions, NbTi bus bars and HTS current leads. We propose three different bus bar designs – for the TF, CS and PF coil feeders, respectively. Two additional values of the TF operating current are considered. The structural verification of the bus bars, including an S-bend geometry for reducing the mechanical loads due to cool down, is performed both analytically and by means of 3D finite-element simulations. In addition, we present an outline design of the HTS current leads, based on the minimisation of the cryogenic heat load and the simulation of a loss-of-flow accident. Finally, we briefly discuss the main characteristics of the joints, as well as some integration and assembly challenges.

Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger

In the framework of the EU DEMO fusion reactor technology development, a new steam generator consisting in a helical tube bundle is currently under study for the lithium-lead loop of the Dual Coolant Lithium Lead (DCLL) and Water Cooled Lithium Lead (WCLL) breeding blankets. This solution turns out to be very interesting for high power plants, since the helical geometry is very compact and it assures a high thermal power exchanged, taking up the minimum amount of space.In this framework, the ENEA Brasimone Research Centre supports the development of this innovative component by means of experimental activities, involving CIRCE, a large scale pool-type facility using Lead-Bismuth Eutectic as primary fluid and pressurized water as secondary fluid. The main components of a new test section named THETIS are: a Fuel Pin Simulator, acting as primary heating source, a vertical mechanical pump for the primary fluid circulation inside the main pool and a new prototypical helical coil steam generator mock-up, designed to be relevant for the DCLL and WCLL lithium-lead loop, acting as primary heat sink. In such configuration, the facility will be involved in a set of tests aiming at demonstrating the technological feasibility and thermal-hydraulic performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. The experiments will also provide a database for system thermal-hydraulic codes validation.The aim of this paper is to present the main layout of the CIRCE facility, to describe the preliminary design of the test section and the main features of the helical coil steam generator mock-up. Furthermore, a preliminary test analysis carried out by the system thermal hydraulic code RELAP5/Mod3.3 is presented. A numerical 1-D model of the helical coil steam generator has been set-up in order to test the performance of the component from a thermal-hydraulic point of view. An additional 3D CFD analysis was performed for the flow field of the HCSG to assess the secondary flow behaviour in the component and the pressure losses.

Development of a PbLi heat exchanger for EU DEMO fusion reactor: Experimental test and system code assessment

In the framework of the DEMO Work Package Balance of Plant of EUROfusion consortium, ENEA has been involved in experimental and numerical activities related to the development of a prototypical heat exchanger, suitable as steam generator for the lithium-lead loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. For this purpose, an experimental campaign has been executed on the pool-type liquid metal-cooled facility CIRCE at ENEA Brasimone Research Centre.A dedicated test section named HERO has been designed and installed inside the main vessel of CIRCE. The innovative steam generator consists of a tube bundle with seven double walled bayonet tubes, fed with pressurized water. The selected configuration improves the plant safety, reducing the possibility of water-lead/lead-alloy interaction thanks to a double physical separation and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas.A set of tests has been defined to demonstrate technological feasibility and performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. In particular, one of the performed tests is presented and discussed in this paper. The experiment is characterized by a secondary loop feedwater working pressure of 10 MPa and a steam generator inlet temperature of 300 °C. On the primary side, the lead-bismuth eutectic has been used as working fluid with a steam generator inlet temperature of 480 °C. During the test, an experimental sensitivity analysis on the primary coolant mass flow rate has been performed. Furthermore, the results of a post-test analysis realized with two versions of the system thermal-hydraulic code RELAP5 are presented, in order to evaluate their capability in simulating the performances of the component and to support the validation process of the codes for heavy liquid metal applications. The work is concluded presenting a scaling analysis to find the equivalence between LBE and PbLi, recalculating the available experimental data with RELAP5 code using PbLi as working fluid.