Fusion Reactor Systems Research Articles

Gas leak tightness of SiC/SiC composites at elevated temperature

SiC/SiC composite materials are attractive candidates for high heat flux components and blanket of fusion reactor, mainly due to their high temperature properties, radiation damage tolerance and low induced radioactivity. One of the challenges for SiC/SiC application in fusion reactors is to satisfy sufficient gas leak tightness of hydrogen and helium isotopes. Although many efforts have been carried-out, SiC/SiC composites by conventional processes have not been successful to satisfy the requirements, except SiC/SiC composites by NITE-methods.Toward the early realization of SiC/SiC components into fusion reactor systems process development of NITE-process has been continued. Followed to the brief introduction of recently developed DEMO-NITE process, baseline properties and hydrogen and helium gas leak tightness is presented.SiC/SiC claddings with 10mm in diameter and 1mm in wall thickness are tested by gas leak tightness system developed. The leak tightness measurements are done room temperature to 400°C. Excellent gas leak tightness equivalent or superior to Zircaloy claddings for light water fission reactors is confirmed. The excellent gas leak tightness suggests nearly perfect suppression of large gas leak path in DEMO-NITE SiC/SiC.

Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

Design, fabrication, and properties of a continuous carbon-fiber reinforced Sm2O3/polyimide gamma ray/neutron shielding material

The design and fabrication of shielding materials with good heat-resistance and mechanical properties is a major problem in the radiation shielding field. In this paper, based on gamma ray and neutron shielding theory, a continuous carbon-fiber reinforced Sm2O3/polyimide gamma ray/neutron shielding material was fabricated by hot-pressing method. The material's application behavior was subsequently evaluated using neutron shielding, photon shielding, mechanical tensile, and thermogravimetric analysis–differential scanning calorimetry tests. The results show that the tensile strength of the novel shielding material exceeds 200MPa, which makes it of similar strength to aluminum alloy. The material does not undergo crosslinking and decomposition reactions at 300°C and it can be used in such environments for long periods of time. The continuous carbon-fiber reinforced Sm2O3/polyimide material has a good shielding performance with respect to gamma rays and neutrons. The material thus has good prospects for use in fusion reactor system and nuclear waste disposal applications.

Mathematical analysis of reduction of copper oxide pellets by hydrogen using the shrinking core model

Reduction of copper oxide by hydrogen at high temperatures to metallic copper is one of the ways of removing hydrogen gas or its isotopes from its mixture with an inert gas like helium which is the coolant for plasma facing first wall of tritium breeding blanket modules in fusion reactor systems. The kinetics of the reaction of hydrogen with copper oxide was obtained from literature and it was used to model the reduction of a single particle of copper oxide using the well-known shrinking core model along with the pseudo-steady state hypothesis. No controlling regime was assumed a priori and the various interfacial gas concentrations were calculated by an iterative procedure based on the pseudo steady state hypothesis as function of the core radius at any time and for given operating conditions. The time required for complete reaction of the pellet was then calculated by numerical integration. A spherical geometry was considered in this work for the purpose of illustrating the technique, but the method is applicable to any kinetic model and any geometry of the pellets after simple modifications to the governing equations.

On the implementation of new technology modules for fusion reactor systems codes

In the frame of the pre-conceptual design of the next generation fusion power plant (DEMO), systems codes are being used from nearly 20 years. In such computational tools the main reactor components (e.g. plasma, blanket, magnets, etc.) are integrated in a unique computational algorithm and simulated by means of rather simplified mathematical models (e.g. steady state and zero dimensional models). The systems code tries to identify the main design parameters (e.g. major radius, net electrical power, toroidal field) and to make the reactor's requirements and constraints to be simultaneously accomplished.In fusion applications, requirements and constraints can be either of physics or technology kind. Concerning the latest category, at Karlsruhe Institute of Technology a new modelling activity has been recently launched aiming to develop improved models focusing on the main technology areas, such as neutronics, thermal-hydraulics, electromagnetics, structural mechanics, fuel cycle and vacuum systems. These activities started by developing: (1) a geometry model for the definition of poloidal profiles for the main reactors components, (2) a blanket model based on neutronics analyses and (3) a toroidal field coil model based on electromagnetic analysis, firstly focusing on the stresses calculations. The objective of this paper is therefore to give a short outline of these models.

Modeling an unmitigated thermal quench event in a large field magnet in a DEMO reactor

The superconducting magnet systems of future fusion reactors, such as a demonstration power plant (DEMO), will produce magnetic field energies in the 10s of GJ range. The release of this energy during a fault condition could produce arcs that can damage the magnets of these systems. The public safety consequences of such events must be explored for a DEMO reactor because the magnets are located near the DEMO's primary radioactive confinement barrier, the reactor's vacuum vessel (VV). Great care will be taken in the design of DEMO's magnet systems to detect and provide a rapid field energy dump to avoid any accidents conditions. During an event when a fault condition proceeds undetected, the potential of producing melting of the magnet exists. If molten material from the magnet impinges on the walls of the VV, these walls could fail, resulting in a pathway for release of radioactive material from the VV. A model is under development at Idaho National Laboratory (INL) called MAGARC to investigate the consequences of this accident in a large toroidal field (TF) coil. Recent improvements to this model are described in this paper, along with predictions for a DEMO relevant event in a toroidal field magnet.

Simulations of fast-wave current drive in pulsed and steady-state DEMO designs

Electromagnetic waves in the ion-cyclotron (IC) range of frequencies are presently investigated as possible current drive (CD) systems in fusion reactors. Among many physical and technical issues, an accurate description of radio-frequency (RF) power absorption by fusion- born alpha particles is of special importance, since RF heating of these particles is not only detrimental for the CD efficiency, but might worsen the operative conditions by increasing their prompt losses.The capability of the full-wave TORIC code has been recently augmented to account for RF absorption by fusion-born alpha particles, calculated to all-orders in finite Larmor radius and with a realistic distribution function. Here, we present simulation with TORIC addressing the sensitivity of current drive efficiency on the design of a future reactor, in particular density and temperature profiles, magnetic field intensity, and plasma dimensions. For this purpose, we have investigated possible frequency windows for CD for two proposed versions of the DEMO reactor, namely its pulsed and its more ambitious steady-state design. The important role of the antenna for a realistic estimate of the CD efficiency is pointed out.

Present status of vanadium alloys for fusion applications

Vanadium alloys are advanced options for low activation structural materials. After more than two decades of research, V–4Cr–4Ti has been emerged as the leading candidate, and technological progress has been made in reducing the number of critical issues for application of vanadium alloys to fusion reactors. Notable progress has been made in fabricating alloy products and weld joints without degradation of properties. Various efforts are also being made to improve high temperature strength and creep-rupture resistance, low temperature ductility after irradiation, and corrosion resistance in blanket conditions.Future research should focus on clarifying remaining uncertainty in the operating temperature window of V–4Cr–4Ti for application to near to middle term fusion blanket systems, and on further exploration of advanced materials for improved performance for longer-term fusion reactor systems.

Preparation and investigation of aluminized coating and subsequent heat treatment on 9Cr–1Mo Grade 91 steel

Iron aluminide inner coating with alumina top layer is being considered as a potential solution for tritium permeation barrier and mitigating MHD pressure drop for liquid metal blanket concepts in the fusion reactor systems. Hot-dip aluminizing with subsequent heat treatment seems to offer a good possibility to produce aluminized coating with alumina top layer. 9Cr–1Mo Grade 91 steel samples were hot dipped in Al melt containing 2.25wt% of Si at 750°C for 3min. Heat treatment was performed at 650, 750 and 950°C for 5h; samples were either air cooled or furnace cooled. Coatings have been evaluated by SEM, EDX, X-ray diffraction, microhardness, scratch adhesion and Raman spectroscopy. The thickness of the layers and phases formed were influenced by the heat treatment adopted. Fe2Al5 was the major phase present in the samples heat treated at 650/750°C, whereas FeAl and α-Fe(Al) primarily made up the outer and inner layers respectively in the samples heat treated at 950°C. Cooling method deployed affected the hardness. Air cooled samples had comparatively higher hardness than furnace cooled samples. The scratch test showed the adhesion for the samples heat treated at 950°C was much better as compared to the samples heat treated at 650/750°C. Raman spectroscopy analysis showed the presence of both α-Al2O3 and γ-Al2O3 on the surface of the samples heat treated at 950°C, while Fe3O4 was present in the furnace cooled sample only.

Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory.Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper.For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available.Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE.In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.

Thermodynamics Properties of Molten Salt Technology Assessment for New Generation Fusion Reactors

In this study, some important thermodynamic properties of the fusion reactor have been analyzed. The physical and chemical properties of molten salts have been extensively studied in the nuclear fusion program. In recent years, molten salts technology began to be used in some engineering areas, in the advanced nuclear field and especially in nuclear fusion reactor systems. Nowadays, Aries team has developed advanced designs by using the molten salts technology in order to get high thermodynamic and structural advantage on nuclear technology areas (Tillack et al. in Fusion Energ Des 65:215–261, 2003; Tillack et al. in Fusion Energ Des 49–50:689–695, 2000; El-Guebaly et al. in Fusion Energ Des 65:263–284, 2003). The Aries-St reactors are a 1000 MW fusion reactor system that based on a low aspect ratio ST plasma (Tillack et al. in Fusion Energy Des 65:215–261, 2003; Tillack et al. in Fusion Energy Des 49–50:689–695, 2000). The Aries team studies especially on liquid walls concepts and this liquid are used to increase neutronic performance of various structures of Aries-St reactors. In this research, candidate molten salts have been studied neutron effects on reactor performance which are the first wall (FW) and blanket. There are various candidate liquids that meet all the criteria such as Li17Pb83, flibe(Li2BeF4) and LiNaBeF4, LiSn that are able to breed enough tritium. In this research, we used Li17Pb83, pure lithium and flibe as candidates that are in the Aries design. Montecarlo n-particle 4b-code is used for neutronics analysis and thermodynamic features. The value of tritium breeding ratio of the Aries-St reactors must be (TBR ≥ 1.1). This can be achieved in the region of LiPb/FW blanket of reactors. Aries-St spherical reactor has high heat flux (0.8 MW/m2) and NWL (6–8 MW/m2) in this region.

Hydrogen and water vapor adsorption properties on cation-exchanged mordenite for use to a tritium recovery system

Tritium recovery system using adsorption or catalytic isotope exchange has already been proposed for a solid breeding blanket system of a nuclear fusion reactor. Synthetic zeolite is often used as an adsorbent or a substrate of chemical exchange catalyst. And, it is well known that its property is changed easily by exchanging its cations. Synthetic mordenite is one of zeolites having fairly large silica/alumina ratio. There are many reports about hydrogen adsorption properties of cation-exchanged mordenite so far. And, the present authors also have reported that cation-exchanged mordenite with Ca ion (Ca–MOR) indicated fairly large hydrogen adsorption capacity at 77K in comparison with Molecular Sieves 5A (MS5A). So, in this work, hydrogen adsorption properties of cation-exchanged mordenite with transition metal ion were investigated mainly. The cation-exchanged mordenite with Ag ion (Ag–MOR) has indicated considerably large hydrogen adsorption capacity in lower pressure range at 77K in comparison with Ca–MOR. The discussion from the viewpoint of adsorption rate is still remaining, but more compact cryosorption column for tritium recovery system is possible to design if Ag–MOR is adopted.

A Primary Energy Conversion System and Design Analysis of a Tokamak Experimental Fusion Reactor

Major efforts are underway to define the objectives of a Tokamak Experimental Fusion Reactor (TEFR). A tokamak is a toroidal chamber that uses a strong toroidal magnetic field to contain high temperature plasma within the torus. Charged particles cannot easily move across strong magnetic fields, and if the fields are closed into nested surface, then deuterium and tritium ions trapped in this way and colliding with sufficient energy to overcome their repulsive Coulomb potential, will fuse and liberate energy. The ultimate goal of this study is to establish the scientific and engineering basis for a detailed reactor design. This paper will concentrate on the TEFR primary energy conversion system (PECS). The PECS includes all the components that lie between the plasma and the toroidal field coil. 

Modeling of MARTe-Based Real-Time Applications With SysML

Model-driven design is recently gaining a wide spreading in different fields, such as design of mechatronics and embedded systems. Approaches based on either UML or SysML permit to efficiently manage the design of complex systems in different areas. In this paper a SysML modeling approach is proposed for the design of real-time applications based on the MARTe framework. The MARTe framework has been recently adopted for the development of real-time systems in several European experimental fusion reactors. The proposed approach allows to achieve a better standardization of the development cycle and a better documentation of the developed systems. Furthermore, by using model-2-text tools it is possible to automatically generate part of the real-time executable code and the application configuration file. The proposed approach has been applied for the modeling of the control system for the Frascati Tokamak Upgrade.

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49–50:689–695, 2000; Tillack et al. in Fusion Eng Des 65:215–261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794–1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3–23, 2006).

Multimodal options for materials research to advance the basis for fusion energy in the ITER era

Well-coordinated international fusion materials research on multiple fundamental feasibility issues can serve an important role during the next ten years. Due to differences in national timelines and fusion device concepts, a parallel-track (multimodal) approach is currently being used for developing fusion energy. An overview is given of the current state-of-the-art of major candidate materials systems for next-step fusion reactors, including a summary of existing knowledge regarding operating temperature and neutron irradiation fluence limits due to high-temperature strength and radiation damage considerations, coolant compatibility information, and current industrial manufacturing capabilities. There are two inter-related overarching objectives of fusion materials research to be performed in the next decade: (1) understanding materials science phenomena in the demanding DT fusion energy environment, and (2) application of this knowledge to develop and qualify materials to provide the basis for next-step facility construction authorization by funding agencies and public safety licensing authorities. The critical issues and prospects for development of high-performance fusion materials are discussed along with recent research results and planned activities of the international materials research community.

Neutronic predesign tool for fusion power reactors system assessment

SYCOMORE, a fusion reactor system code based on a modular approach, is under development at CEA. In this framework, this paper describes a methodology developed to build the neutronic module of SYCOMORE. This neutronic module aims to evaluate main neutronic parameters characterising a fusion reactor (tokamak): tritium breeding ratio, multiplication factor, nuclear heating as a function of the reactor main geometrical parameters (major radius, elongation, etc.), of the radial build, Li enrichment, blanket and shield thickness, etc. It is based on calculations carried out with APOLLO2 and TRIPOLI-4 CEA transport code on simplified 1D and 2D neutronic models. These models are validated versus a more detailed 3D Monte-Carlo model (using TRIPOLI-4). To ease the integration of this neutronic module in SYCOMORE and provide results instantly, a surrogate model that replicates the 1D and 2D neutronic model results was used. Among the different surrogate models types (polynomial interpolation, responses functions, interpolating by Kriging, artificial neural network, etc.) the neural networks were selected for their efficiency and flexibility.The methodology described in this paper to build SYCOMORE neutronic module is devoted to HCLL blanket, but it could be applied to any breeder blanket concept provided that appropriate validation could be carried out.

Opportunities for Advanced Ceramics and Composites in the Nuclear Sector

Ceramics have played a crucial role in the development of fission based nuclear power, in glass & glass composite high level wasteforms, in composite cements to encapsulate intermediate level wastes (ILW) and also for oxide nuclear fuels based on UO2 and PuO2/UO2 mixed oxides. They are also used as porous filters with the ability to absorb radionuclides (RN) from air and liquids and are playing a key role in the cleanup at Fukushima. Non‐oxides also find current fission applications including in graphite moderators and B4C control rods. Ceramics will continue to be significant in the near‐term expansion of nuclear power via next‐step developments of fuels with inert matrices or based on thoria and in wasteforms using alternative composite cements or single or multiphase ceramics that can host Pu & other difficult RN. Longer term advances for Generation IV reactors, which will operate at higher temperatures & with higher fuel burn‐up require innovative fuel developments potentially via carbides & nitrides or composite fuel systems. Novel non‐thermal (cement‐like) and thermal techniques are currently being developed to treat some of the difficult legacy wastes. Non‐thermally derived wasteforms developed from geopolymers, composite cements, hydroceramics, and phosphate‐bonded ceramics and thermally derived wasteforms made by Hot Isostatic Pressing and fluidized bed steam reforming (FBSR) as well as vitrification techniques based on cold crucible melting (CCM), Joule‐heater in‐container melting and plasma melting (PM) are described. Future developments in waste treatment will be based on separation technologies for partitioning individual RN along with design & construction of RN‐containing ceramic targets for inducing transmutation reactions. Near demonstration actinide‐hosting ceramic wasteforms including multiphase Synroc systems are described. Opportunities also exist for ceramics in structural applications in Generation IV reactors such as composite SiC/SiC and C/C for fuel cladding and control rods and MAX phases and ultrahigh‐temperature ceramics (UHTCs) may find near core fuel coating and cladding applications. Uses of ceramics in fusion reactor systems will be both functional (ceramic superconductors in magnet systems for plasma control and in Li silicate breeder blankets in tokamaks) and structural including as sapphire diagnostic windows, graphite diverters, and plasma facing C and UHTCs. In all these cases, performance is limited by poorly understood radiation damage and interface controlled processes, which demands a combined modeling/experimental approach.

Recent research activities on functional ceramics for insulator, breeder and optical sensing systems in fusion reactors

The paper presents a brief overview of current research activities on functional ceramic materials for insulating components, tritium breeder and optical sensing systems, mainly carried out at Institute for Materials Research (IMR), Tohoku University. Topics include recent experimental results related to the electrical degradation and optical changes in typical oxide ceramics (e.g. Al2O3 and SiO2) concerning radiolytic effects. Hydrogen effects on the electrical conductivity in the Perovskite-type oxide ceramics and the interaction between hydrogen and irradiation induced defects in ternary Li oxides used as breeder materials, were dynamically observed under the irradiation environment. Further attention is focused on several challenging qualifications required for an advanced sensing system using optical characteristics (e.g., thermoluminescence in SiO2 core fiber, neutron-induced long lasting emission from oxides doped with rare-earth elements, and gasochromic coloration phenomenon of WO3).

Effect of thermal cycling of SiCf/SiC composites on their mechanical properties

SiCf/SiC composites are class of high temperature structural materials being developed for use in nuclear fusion and fission reactor systems because of their superior high temperature mechanical properties, low radiation damage and low induced radioactivity. Two types of 2D SiCf/SiC composites were made through isothermal and isobaric chemical vapor infiltration process using eight harness satin-woven ceramic-grades Nicalon™ fibers with boron nitride (BN) interface, namely: one with lower interface thickness and a second type with higher interface thickness. The BN interface was applied to the fiber prior to SiC matrix addition to modify the interfacial bond strength leading to better toughness and improved oxidation resistance. The density achieved was around 2.6g/cc. The composite specimens were subjected to thermal cycling treatment using an in-house furnace. The mechanical properties such as tensile strength, fracture toughness and interfacial bond strength were also studied for all the composites before and after thermal cycling. It is seen from the results that both composites withstood thermal shocks and thermal cycling treatment. It was also concluded from the present work that good balance between load transfer and crack arresting was established.