Nuclear Fusion Plant Research Articles

Accuracy and scalability of incompressible inductionless MHD codes applied to fusion technologies

Abstract It is well-known that magnetohydrodynamics (MHD) dominates the dynamic of the liquid metal flows inside the breeding blankets (BB) of future nuclear fusion plants by magnetic confinement. MHD is a multiphysics phenomenon involving both electromagnetism and incompressible fluid mechanics. From the computational point of view, the simulation of MHD flows in fusion relevant conditions entails a significant challenge. Indeed, due to the shape of the induced electrical currents inside the bulk of the fluid, high spatial resolutions are needed to capture the large gradients found in boundary layers and 3D effects. Besides, solving the equations accurately typically requires very small time steps for the transient algorithms. Over the past few decades, some parallel MHD codes have been developed with success to simulate complex flows in increasingly realistic geometries. Among them, the MHD tools of commercial CFD platforms have attracted attention due to their relatively soft learning curve. Most of these codes are based on the so called ϕ-formulation which, by applying the divergence free condition of the current density to the Ohms law, reduces the electromagnetic part of the problem to a single Poisson equation. As a downside, the approach segregates the fluid and electromagnetic problem. In practice, this establishes important limits to the mesh element size, to the mesh quality and to the time-step needed to obtain accurate and stable solutions that maintains charge conservation at a discrete level. In this work, these limits are explored for the commercial platform ANSYS-Fluent using a test geometry under different conditions. As an alternative, a new code based on Finite Element Methods (FEM) is introduced as well. This open-source code, called GridapMHD (https://github.com/gridapapps/GridapMHD.jl), aims at solving the full set of MHD equations using a monolithic approach. GridapMHD is still in early stages of development but it has already shown promising results.

Experimental Characterization of an NEG Pump of Novel Size—A Major Step toward Its Application in DEMO Neutral Beam Injectors

A future nuclear fusion plant DEMO needs neutral beam injection (NBI) systems requiring a vacuum pumping system with a very high pumping speed, in the order of several 1000 m3/s. Large customized cryopumps are actually used to meet the requirements. A promising concept for future NBI systems is based on high capacity getter materials. The ZAO® alloy, developed by SAES Getters, Italy, provides a drastically improved performance for the pumping of hydrogen compared to conventional getter materials. This paper describes the experimental characterization of a large pump of scalable size with 15 kg of ZAO® and the achieved results, in particular the systematic investigation of sorption characteristics and regeneration behaviors. Major findings include a very good repeatability of the sorption performance, a reduced pumping speed at higher pressures only above the NBI relevant level, an improved performance (+20%) with elevated getter temperature and an isotope independent sticking coefficient for hydrogen. Furthermore, improved operation experience and regeneration prediction tools have been developed. Employing the experimental results, a simulation task was performed and the sticking factor of the getter cartridge was determined with 7%.

Passive Hydrogen Recombination during a Beyond Design Basis Accident in a Fusion DEMO Plant

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO.

Dynamic Event Tree Analysis as a Tool for Risk Assessment in Nuclear Fusion Plants Using RAVEN and MELCOR

In the broad framework of the nuclear power plants (NPPs) industry, the dynamic probabilistic risk assessment could answer the time dependence deficiency of the event tree and fault tree analysis. The basic event tree approach relies on experts’ pre-constructed accident sequences without exploring the time-dependent nature of an accident scenario, which could strongly affect the accident sequence. Conversely, the effects of events timing can be studied by adopting a dynamic event tree (DET) approach. Developing a DET methodology requires integrating a system code capable of replicating an accident scenario and a logic-driver code able to generate the event tree sequence, trigger plant safety systems, and manage other relevant events throughout the simulation. For this purpose, Methods for Estimation of Leakages and Consequences of Release (MELCOR) and reactor analysis and virtual control environment (RAVEN) have been coupled through a Python script developed by the Sapienza University of Rome, Rome, Italy, to perform DET studies during accident transients in fusion and fission reactors. RAVEN is a software tool developed at the Idaho National Laboratory (INL), Idaho Falls, ID, USA, to act as a control logic driver and post-processing tool for different applications. MELCOR for fusion is a fully integrated design basis and severe accident code that simulates thermal-hydraulic behavior and self-consistently accounting for aerosol transport in nuclear facilities and reactor cooling systems for the evaluation of the source term in fusion reactors. The coupling between these codes will provide a wide range of NPP risk assessment analyses, establishing new best practices. In this work, a preliminary DET study has been performed, selecting as initiating event an ex-vessel loss of coolant accident (LOCA) in the water-cooled lithium lead test blanket system (WCLL TBS) to be tested in International Thermonuclear Experimental Reactor (ITER). Time-dependent parameters such as the intervention of the plasma shutdown system and the closure of the main system isolation valves have been sampled to study evolving system scenarios.

Overview on the management of radioactive waste from fusion facilities: ITER, demonstration machines and power plants

In the absence of official standards and guidelines for nuclear fusion plants, fusion designers adopted, as far as possible, well-established standards for fission-based nuclear power plants (NPPs). This often implies interpretation and/or extrapolation, due to differences in structures, systems and components, materials, safety mitigation systems, risks, etc. This approach could result in the consideration of overconservative measures that might lead to an increase in cost and complexity with limited or negligible improvements. One important topic is the generation of radioactive waste in fusion power plants. Fusion waste is significantly different to fission NPP waste, i.e. the quantity of fusion waste is much larger. However, it mostly comprises low-level waste (LLW) and intermediate level waste (ILW). Notably, the waste does not contain many long-lived isotopes, mainly tritium and other activation isotopes but no-transuranic elements. An important benefit of fusion employing reduced-activation materials is the lower decay heat removal and rapid radioactivity decay overall. The dominant fusion wastes are primarily composed of structural materials, such as different types of steel, including reduced activation ferritic martensitic steels, such as EUROFER97 and F82H, AISI 316L, bainitic, and JK2LB. The relevant long-lived radioisotopes come from alloying elements, such as niobium, molybdenum, nickel, carbon, nitrogen, copper and aluminum and also from uncontrolled impurities (of the same elements, but also, e.g. of potassium and cobalt). After irradiation, these isotopes might preclude disposal in LLW repositories. Fusion power should be able to avoid creating high-level waste, while the volume of fusion ILW and LLW will be significant, both in terms of pure volume and volume per unit of electricity produced. Thus, efforts to recycle and clear are essential to support fusion deployment, reclaim resources (through less ore mining) and minimize the radwaste burden for future generations.

Triple ion beam irradiation of glass-ceramic materials for nuclear fusion technology

The analysis of the radiation damage associated with energetic neutrons is a key factor for the application of glass-ceramics as joining materials in nuclear fusion plants. Triple ion beam irradiation is, currently, the only way to emulate the displacement damage and the effects of transmutation products induced by high energy neutrons at elevated doses by producing a high level of displacements per atom (dpa) and implanting high concentrations of low solubility gases on the glass-ceramic microstructure. In this paper, we investigate calcia-alumina glass-ceramics in which the damage expected in the working conditions of a nuclear fusion plant were reproduced by simultaneous irradiation with Si, He and H ions at different temperatures. TEM analysis of irradiated calcia-alumina samples shows amorphization (due to the Si beam) and partial recrystallization of the irradiated portion and the formation of He bubbles/cavities. The irradiation temperature plays a prominent role in determining the induced damage in terms of recrystallization, phase mixing and bubble size, density and aggregation.

DEMO – The main achievements of the Pre – Concept phase of the safety and environmental work package and the development of the GSSR

The EUROfusion Safety and Environmental Work Package (WPSAE) has the scope to progress the safety studies for the future EU DEMO reactor. A Generic Site Safety Report (GSSR) has been prepared to include all the steps necessary to cover the safety issues of the nuclear fusion plant, from the definition of principles and requirements, through the detection of the source terms at risk, selecting the postulated initiating events, analyzing the accidents, quantifying the doses to the population and investigating waste production and its management. Eleven GSSR volumes collate the studies performed. The final goal is to prepare safety documentation, as complete as possible, to initiate a preliminary safety report when the plant site is selected.In parallel, the safety studies are supporting the maturation of the design of DEMO, providing feedback on the technical choices for the machine, the selection of materials, the use of space, the equipment necessary to correctly manage the safety risks through continuous collaboration with the design teams of other DEMO Work Packages.The GSSR focused on safety and environmental issues, it did not encompass security issues that could impact the design and hence the safety performance of DEMO. Consideration of the interactions between safety and security will be needed in the development of the DEMO detailed design and its associate Preliminary Safety Analysis Report (PSAR).The main achievements of the Pre-Concept Phase are presented in this paper, by summarizing the contents of GSSR volumes. Completion of the work is foreseen during the future Concept Design Phase (2021–2027).

Preliminary Assessment of Cooling Water Chemistry for Fusion Power Plants

The determination of the water chemistry for cooling systems of nuclear fusion plants is under debate. It should be tailored for different types of fusion reactors: either experimental, e.g., ITER, JT-60SA, and DTT, or aimed at power generation, e.g., DEMO, given the different operation requirements. This paper presents the dual approach involving experiments and computer simulations chosen for the definition of DEMO water chemistry. Experimental work was performed to assess the corrosion susceptibility of reduced activation ferritic martensitic EUROFER 97 and AISI 316L in different water chemistry regimes. At the same time, the low corrosivity requirement brings an additional safety aspect for the radiation protection since some neutron-activated corrosion products (ACPs) create a gamma radiation when deposited outside the plasma chamber in components accessible to operators and these must be minimized. To evaluate the ACP inventory for DEMO, assessments were carried out using a reference computer code. Preliminary experimental activities to define the water chemistry of DTT under construction at ENEA were also conducted. The comparison of code results with experiments is two-fold important: for the validation of the computer code models and to determine data that are necessary to perform calculations.

Preliminary Design of the Electrical Power Systems for DTT Nuclear Fusion Plant

The realization of the Divertor Tokamak Test (DTT) facility is one of the key milestones of the European Roadmap, aiming to explore alternative power exhaust solutions for DEMO, the first nuclear-fusion power plant that will be connected to the European grid. For the actual implementation of the DTT and DEMO plants, it is necessary to define the structure of the internal electric power distribution system, able to supply unconventional loads with a sufficient level of reliability. The present paper reports the preliminary studies for the feasibility and realization of the electrical power systems of DTT, describing the methodology adopted to obtain a first distribution configuration and providing some simulation results. In particular, the first stage of the study deals with the survey and characterization of the electrical loads, which allows defining a general layout of the facility and size the main electrical components. To verify the correctness of the assumptions, simulation models of the grid were implemented in the DIgSILENT PowerFactory software in order to carry out power flow and fault analyses.

Analysis of nodalization strategies to model a suppression tank for fusion plants with TRACE code

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the main safety issues in nuclear fusion plants. The ICE is caused by the rupture of the coolant tubes installed in the plasma facing components; the coolant ingress causes a pressure rise in the plasma chamber and vacuum vessel, normally under high-vacuum conditions. To mitigate the system pressurization leading to mechanical structure failure, a pressure suppression system is installed. Safety analyses of the hypothetical challenging accidental scenarios can be conducted by deterministic models that need to be validated against experimental data characterizing the target phenomena. The paper presents a study of different nodalization strategies for modeling a suppression tank for fusion plants, using the best estimate thermal-hydraulic system code TRACE (TRAC/RELAP Advanced Computational Engine). The TRACE code is developed by USNRC to perform safety analyses for light water fission reactors. Both mono-dimensional and three-dimensional approaches were adopted to model the suppression tank. The experimental data from the JAERI Integrated ICE facility (scaling factor of 1/1600 with respect to the ITER-FEAT design) was used to benchmark different nodalization options. In addition to the qualitative accuracy evaluation, the Fast Fourier Transform Based Method (FFTBM) was applied to quantify the accuracy of the code results for different nodalizations.

Ingress of Coolant Event simulation with TRACE code with accuracy evaluation and coupled DAKOTA Uncertainty Analysis

Among the Postulated Initiating Events in nuclear fusion plants, the Ingress of Coolant Event (ICE) in the Plasma Chamber is one of the main safety issues. In the present paper, the best estimate thermal-hydraulic system code TRACE, developed by USNRC, has been adopted to study the ICE, and it has been qualified based on experimental results obtained in the Integrated ICE facility at JAERI. A nodalization has been developed in the SNAP environment/architecture, using also the TRACE 3D Vessel component where multidimensional phenomena could occur. The accuracy of the code calculation has been assessed both from a qualitative and quantitative point of view. In addition, an Uncertainty Analysis (UA), with the probabilistic method to propagate the input uncertainties, has been performed to characterize the dispersion of the results. The analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in the SNAP environment/architecture. Results show the adequacy of the 3D nodalization and the capability of the code to follow the transient evolution also at a very low pressure. Response correlations have been computed to characterize the correlation between the selected uncertain input parameters and the Plasma Chamber pressure.

Numerical analysis of sub-atmospheric steam condensation in suppression tank with SIMMER IV code

Abstract One of the key safety components for nuclear fusion plants is the suppression tank, which is designed to protect the Vacuum Vessel (VV) against accidental pressurization events, e.g. Loss Of Coolant Accident (LOCA). In this framework the attention is focused on the Vacuum Vessel Pressure Suppression System (VVPSS), made of water tanks in which the pure steam, or eventually mixed with incondensable gases, is injected and consequently the overpressure is dumped profiting of Direct Contact Condensation (DCC). The design constraints of fusion reactor dictate that the pressure resulting (long-term) from any accidental or baking condition should be always kept lower than 0.15 MPa. The study of the phenomena evolving during DCC in LOCA conditions is the major novelty, especially in consideration of the lack of similar studies in the available literature. In this context, a wide series of experimental tests was carried out at Pisa University (UNIPI), Department of Civil and Industrial Engineering (DICI), in a Small Scale Test Facility (SSTF), designed and instrumented for investigating DCC at sub-atmospheric pressure, by varying water pool temperature, pressure and steam mass flow rate. The adoption and assessment of suitable numerical codes, to reliably simulate such a cutting-edge multiphase multicomponent scenario, have a crucial role for contributing to the phenomena understanding and for possible safety analysis of full-scale components. On this basis, a preliminary evaluation of the cartesian three-dimensional SIMMER IV code capabilities in simulating DCC at sub-atmospheric conditions was carried out, taking as reference one UNIPI test. SIMMER IV code was able to set up precise initial low-pressure boundary conditions and simulate superheated steam condensation in subcooled water pool, with condensation efficiency comparable to the experimental one. Moreover, SIMMER IV code predicted a longitudinal steam plume dimension and injected steam velocity consistent with experimental data.

Exploration of power conversion thermodynamic cycles for ARC fusion reactor

Abstract In the worldwide energy industry, nuclear fusion could be a breakthrough in the medium-long term. One promising fusion machine under design at Massachusetts Institute of Technology is ARC reactor. It is likely that the first nuclear fusion plants will rely on a traditional thermodynamic cycle for the downstream power energy conversion. In this framework, one of the design aspects is to maximize the thermal efficiency. In the present paper the thermodynamic cycles, which could be adopted in ARC rector, are explored. Three cycles have been considered: the Rankine, the Brayton and a combined cycle. For the gas adopted in the Brayton and combined cycles, two options have been investigated: supercritical Helium and supercritical CO2. A comparison among thermal efficiency and preliminary considerations on component integrity’s, plant feasibility and economics of each studied configurations has been discussed to identify the possible best option for ARC reactor. The results show that a regenerative CO2 Brayton cycle with intercooler and re-heating systems is the most promising one. Such configurations is able to reach a thermodynamic efficiency of up to 0.6.

Radiation Protection Design and Licensing for an Experimental Fusion Facility: The Italian and European Approaches

An experimental nuclear fusion device could be seen as a step toward the development of the future nuclear fusion power plant. If compared with other possible solutions to the energy problem, nuclear fusion has advantages that ensure sustainability and security. In particular, considering the radioactivity and the radioactive waste produced in a nuclear fusion plant, the component materials for the plant could be selected in order to limit the decay period, making recycling possible in a new reactor after about 100 yr from the beginning of decommissioning. To achieve this and other pertinent goals, many experimental machines have been developed and operated worldwide in the last decades, underlining that radiation protection and worker exposure are critical aspects of these facilities due to the high-flux, high-energy neutrons produced in the fusion reactions. Direct radiation, material activation, tritium diffusion, and other related issues pose a real challenge to demonstrating that these devices are safer than nuclear fission facilities. In Italy, for the past 30 yr, a limited number of fusion facilities have been constructed and operated, mainly at the ENEA Frascati Center, where a new one, the Italian Divertor Tokamak Test Facility (DTT), is now under development. The radiation protection approach, addressed by national licensing requirements, shows that respecting the constraints for worker exposure to ionizing radiation is not always straightforward. In the current analysis the main radiation protection issues encountered in the Italian fusion facilities are considered and discussed, and the technical and legal requirements are described. The licensing process for this kind of device is outlined and compared with that of other European countries.The following aspects are considered throughout the current study: description of the installation, plant, and systems; suitability of the area; buildings and structures; radioprotection structures and organization; exposure of personnel; accident analysis and relevant radiological consequences; and radioactive waste assessment and management.In conclusion, the analysis points out the need for special attention to the radiological exposure of workers in order to demonstrate at least the same level of safety as that reached at nuclear fission facilities.

Effect of cooling rates on solidification, microstructure and mechanical properties in tungsten

Tungsten with its excellent high-temperature properties would be a most promising candidate as a plasma-facing material at the divertor in a nuclear fusion plant.

Meso-scale modeling and simulation for reduced activation ferritic/martensitic steel

Reduced activation ferritic martensitic (RAFM) steel has been regarded as a candidate for blanket material of nuclear fusion plant. Microstructure of RAFM steel, featured by precipitates and hierarchical structure of martensite, contribute to its excellent irradiation resistance. For better understanding of mechanical responses of RAFM steels, microstructure-based simulation has been developed for describing detailed elastic-plastic behavior in grain (or meso-) scale. In this study, a dislocation density based crystal plasticity model including dislocation-precipitate interactions was proposed and implemented in the finite element simulations of meso-scale micro-pillar compression and macro-scale uniaxial tensile tests. For the macroscale simulations, a representative volume element approach with periodic boundary condition was applied for predicting mechanical response of investigated polycrystalline RAFM steel based on measured mesoscale parameters and microstructure. The presented crystal plasticity model could reproduce the mesoscopic local deformation and macroscale flow stress strength in good agreement.

Temperature Model Identification of FTU Liquid Lithium Limiter

In this brief, the thermal model identification of the limiter surface placed in nuclear fusion plants facilities is presented. Experimental data measured at the Frascati Tokamak Upgrade (FTU) have been considered in order to identify a nonlinear dynamical model of the temperature distribution over the cooled lithium limiter used in the FTU. Three data-driven approaches have been followed: 1) a linear autoregressive (ARX) model; 2) a nonlinear ARX model; and 3) a Hammerstein model. The selection procedure of the relevant physical quantities will be outlined and a comparison among the obtained models will be given.

Model identification of the temperature over the FTU cooled liquid lithium (CLL) limiter surface

In this work, a data-driven model procedure aimed at identifying the dynamical behavior of the thermal distribution over the limiter surface placed in nuclear fusion plant facilities is provided. In particular, we focused on experimental data collected in the in Frascati Tokamak Upgrade (FTU) where a cooled liquid lithium limiter (CLL) with a capillary porous system (CPS) technology is adopted. Thus, a nonlinear dynamical system, namely a Hammerstein–Wiener model, has been investigated and opportunely identified in order to predict the thermal distribution over the CLL. The identified system is able to predict spatiotemporal thermal information even in the worst case scenario when disruptive events occur during plasma experiments.

Hydrogen interaction characteristic of nanoscale oxide films grown on iron–nickel based stainless steel by selective thermal oxidation

Hydrogen isotopes, the reaction ingredients in the nuclear fusion plant, can easily permeate through the stainless steel (SS) substrate, leading to the so-called hydrogen degradation. Generally, a widely accepted way to reduce the hydrogen permeation is to prepare a barrier coating on the substrate. Nevertheless, the coated layer has the inherent problem of incompatibility with the heterogeneous base materials. In this work, in-situ selective oxidation was used to explore the optimal oxides with the improved hydrogen resistance. Two types of layers thermally formed at 450 °C and 750 °C, respectively, were selected to investigate their hydrogen interaction characteristics. Comprehensive analyses, including Raman spectra, XPS, EIS and AES, indicate that the oxide formed at 450 °C is a better candidate of hydrogen permeation barriers, probably due to the formation of protective layers of chromia and FeCr2O4, while the oxides obtained at 750 °C, though exhibiting a much more stable phase, can rarely reduce hydrogen diffusion through the shortcuts of defects. This finding provides a potential new way to prepare a hydrogen permeation barrier.

Shutdown dose-rate assessment during the replacement of in-vessel components for a fusion DEMO reactor

The shutdown dose rate at Japan’s demonstration nuclear fusion (DEMO) plant was analyzed using MCNP-5 and DCHAIN-SP2001 to assess its significance in the development of a maintenance program. It was assumed that the blanket segments were integrated with a shielding plug (SP) and replaced through vertical upper ports, whereas divertor cassettes, also integrated with the SP, were replaced through bottom ports. To minimize the dose rate, remote handling equipment should approach from behind the SP and be fixed with an attachment to it. The estimated dose rates during manifold cutting and rewelding in the maintenance ports were as low as 0.01Gy/h for the blanket segment and 0.1Gy/h for the divertor cassette. When the outboard blanket segments and divertor cassettes were removed along with the SP, a spatial dose rate of 100Gy/h was found at the maintenance ports. The spatial dose rate of the in-vacuum vessel at ITER during maintenance using remote handling equipment is limited to 250Gy/h. The study confirmed that a maintenance scheme involving pipe cutting or rewelding would be applicable to ITER at the estimated spatial dose rates.