Overview of the HCPB Research Activities in EUROfusion Hernandez, Francisco A.; Arbeiter, Frederik; Boccaccini, Lorenzo V. ...
IEEE transactions on plasma science,
2018-June, 2018-6-00, Volume:
46, Issue:
6
Journal Article
Peer reviewed
In the framework of the EUROfusion's Power Plant Physics and Technology, the working package breeding blanket (BB) aims at investigating four different BB concepts for an EU demonstration fusion ...reactor (DEMO). One of these concepts is the helium-cooled pebble bed (HCPB) BB, which is based on the use of pebble beds of lithiated ternary compounds and Be or beryllides as tritium breeder and multiplier materials, respectively, EUROFER97 as structural steel and He as coolant. This paper aims at giving an overview of the EU HCPB BB Research and Development (R&D) being developed at KIT, in collaboration with Wigner-RCP, BUTE-INT, and CIEMAT. The paper gives an outline of the HCPB BB design evolution, state-of-the-art basic functionalities, requirements and performances, and the associated R&D activities in the areas of design, functional materials, manufacturing, and testing. In addition, attention is given also to the activities dedicated to the development of heat transfer augmentation techniques for the first wall and the corresponding testing. Due to their nature as design drivers, a brief overview in the R&D of key HCPB interfacing areas is given as well, namely, the tritium extraction and recovery system, the primary heat transfer and power conversion systems, and safety topics, as well as some specific activities regarding the integration of in-vessel systems through the BB. As concluding remarks, an outline of the standing challenges and future R&D plans is summarized.
Demonstration fusion reactor (DEMO) will aim to fulfill the tritium self-sufficiency in fuel recycling and generate economical fusion energy in the future. The tritium breeding blanket is one of the ...key components inside the vacuum vessel of a fusion reactor. Achieving tritium self-sufficiency is one of the most important goals, which involves whether fusion reactors can achieve fuel cycle and sustained stable operation. Furthermore, the tritium release efficiency is also an important factor affecting the operation of DEMO, which is intimately related to the temperature field inside the breeding blanket. Consequently, the research is focused on the thermal principles of thermal hydraulics in the breeding blanket for the tritium generation process. In this work, a preliminary study considering the blanket structure and the cooling conditions based on a PWR water-cooled blanket was carried out. A neutronics model was established using the Monte Carlo N Particle Transport Code (MCNP) and the nuclear heat distribution in the blanket was analyzed. The nuclear heat is used as the preprocessing data for thermal calculation. The ANSYS code is used to perform the finite element technique analysis on the coolant parameters related to the temperature field, such as the inlet velocity, pipe diameter, pipe spacing, and so on. The breeding blanket met the temperature requirements while also possessing ideal tritium release performance. This work will be beneficial to the DEMO blanket design activities and as a useful reference in the future.
A purely geometric approach has been investigated to reconstruct the Demonstration Fusion Power Reactor (DEMO) plasma boundary for control purposes. The whole plasma boundary is reconstructed by ...using a deformable template method based on B-splines. The final curve shape is achieved by minimizing the distance between a limited number of estimated and measured (at present provided by an equilibrium code) plasma boundary points along the reflectometer lines of sight. The resulting unconstrained optimization problem is solved by a simulated annealing algorithm. The method is complemented by including the available plasma and poloidal field coil current measurements to refine the boundary reconstruction in the X-point region. The robustness with respect to random measurement random errors and to a reduction in the number of measurements is discussed. The main equilibrium and shape geometric quantities (such as plasma cross-sectional area, plasma center position, elongation, and triangularity) were computed and compared to the corresponding quantities of a DEMO reference equilibrium.
The reflectometry diagnostic for demonstration fusion reactor (DEMO) is envisaged to provide the electron density profile and to be used as a control diagnostic for the real-time vertical position ...controller. An initial conceptual study has been performed defining the position of the antennas and the routing of the waveguides. In the present design, the integration has been driven by the remote handling and blanket interfaces. To progress with the system integration, neutronics simulations were performed to assess the cooling requirements. This paper presents a nuclear and thermal analysis for the initial design of the reflectometry diagnostic for DEMO. The neutronics simulations were performed using the Monte Carlo simulation program MCNP6 and FENDL 2.1 cross sections, while ANSYS mechanical v18 was used to perform the thermal analysis. Simulation results show that the nuclear heat loads reach 8 W/cm 3 at the surface of the diagnostics section. Without an active cooling system, the operating temperatures of the components under such heat loads would be well above the acceptable from the thermomechanical point of view. The thermal analysis here presented provides the temperature distribution in the components when subjected to neutron/gamma irradiation and thermal radiation from the plasma, after the implementation of a preliminary design of the active cooling system. The operation temperatures of the plasma-facing antennas, as well as the performance of the diagnostics system from the neutron shielding point of view, are analyzed and discussed in this paper.
In the framework of the European fusion program for energy, EUROfusion funds the studies for the future fusion power demonstrator reactor DEMOnstration fusion reactor (DEMO). Commissariat à ...l'energie atomique et aux energies alternatives (CEA) is involved in the conceptual design of the superconducting conductors for the toroidal field (TF) and central solenoid (CS) magnets. The CEA design proposal corresponds to wind and react Nb3Sn pancake-wound coils using cable-in-conduit conductors cooled at about 4.5 K by forced flow of supercritical helium. This paper presents the latest thermal-hydraulic analyses performed on both TF and CS conductors. Two TF conductor designs have been analyzed, with nominal currents of 111 kA and 88 kA, respectively. The analyses were performed in normal (burn) and off-normal (quench) conditions. Burn simulations focused on the central and lateral pancakes with heat load corresponding to neutron heating. The central pancake is the most critical one regarding the magnetic field, while the lateral one receives a more important heat load from the casing. The influence of case cooling on temperature margin (ΔTma) was analyzed by means of a dedicated 2D Cast3M model. The ΔTma sensitivity to driving parameters such as conductor friction factor correlations was analyzed. The impact of interturn thermal coupling on temperature margin was also assessed. Quench studies were performed on the TF conductor, considering a quench initiated either on the innermost turn or at the middle of hydraulic length. Regarding the CS conductor, burn scenario was led on the latest design featuring a nominal current of 53.7 kA, focusing on the impact of ac losses and of dwell duration.
The cryo-magnetic system of the future tokamak EU DEMO includes the superconducting magnets cooled by a forced flow of supercritical helium at about 4.5 K, the cryo-distribution lines and valve ...boxes, and the cryogenic system with several cold boxes. The DEMO cryogenic system is at the pre-conceptional design phase, with the identification of the heat loads requirements of the main cryogenic users. The analysis will focus on the cooling capacity of the superconducting magnets, with the application on a toroidal field coil conductor based on a CEA design. Simcryogenics, a dynamic modelling tool developed by CEA, has been used to model a supercritical helium loop for cooling one conductor pancake. An algorithm has been developed to optimize the cold source temperature and the supercritical helium mass flow, in order to minimize the refrigeration power. At this pre-conceptional design phase, the interest of such analysis is to contribute to the optimization of both the cryogenic system and the superconducting magnets to minimize the refrigeration power, one of the important investment and operation costs for the future tokamak.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UILJ, UL, UM, UPCLJ, UPUK, ZAGLJ, ZRSKP
The nuclear physical properties of austenitic chromium-nickel steel EK-164 (Fe–16Cr–19Ni–2Mo–2Mn–Nb–Ti–B) and its manganese-based modifications EK-164Mn (Fe–16Cr–20Mn–2Mo–Nb–Ti–B) and EK-164MnW ...(Fe–16Cr–20Mn–2W–Nb–Ti–B) under irradiation (up to 5 effective years) in the neutron spectra of fast power (BN-600) and thermonuclear (DEMO-C) reactors and after irradiation (nuclear cooling up to 1000 years) were determined. The investigations were based on using the ACDAM-2.0 calculation complex with consideration of the elemental (alloying and impurity) compositions of steels and typical neutron spectra of fission (BN-600) and fusion (DEMO-C) reactors. The entire set of nuclear physical characteristics of steels (primary radiation damage; transmutation of elements; radiogenic accumulation of elements, including hydrogen, helium, and boron; γ-dose rate; radioactivity; nuclear energy release; isotope contributions) is obtained and their comparative analysis is performed. The time intervals for fulfilling the criteria of low activation of steels after irradiation (γ-dose rate less than 10 mSv/h, radioactivity less than 10
10
Bq/kg, energy release less than 10 W/m
3
) were determined.
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DOBA, EMUNI, FIS, FZAB, GEOZS, GIS, IJS, IMTLJ, IZUM, KILJ, KISLJ, MFDPS, NLZOH, NUK, OILJ, PILJ, PNG, SAZU, SBCE, SBJE, SBMB, SBNM, SIK, UILJ, UKNU, UL, UM, UPUK, VKSCE, ZAGLJ
Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and ...fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥1 km/s. An innovative approach, aimed at identifying suitable inboard "direct line" paths, to inject high-speed pellets (in the 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. A possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.
Plasma enhancement gases (PEGs) (such as: nitrogen, neon, argon and other inert gases) are injected into the plasma of several tokamaks in order to reduce the power load over the plasma facing ...components. The exhaust gas in the tokamak demonstration reactor (DEMO) consists of more than 90% of unburned fuel gas (D and T) and the remaining part will be He, PEGs and impurities. In DEMO reactor it is foreseen to recover the fuel gas and PEGs. The research focuses to remove the He from fuel gas and PEGs from fusion reactor. For this purpose, six commercial ceramic membranes have been tested at a permeation apparatus built at ENEA Casaccia laboratories. Single gas permeances for H2, He, Ar and air were measured with a pressure drop across the membranes between 10 Pa up to 100 kPa at room temperature. The selectivities found are low except for the 0.5 nm pore size membranes. The membranes have been supplied by Ceramiques Tecniques et Industrielles (CTI SA, France).
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UILJ, UL, UM, UPCLJ, UPUK, ZAGLJ, ZRSKP
Plasma enhancement gases (PEGs) (such as: N2, Ne, Ar and other inert gases) are injected into the plasma of several tokamaks in order to reduce the power load over the plasma facing components. The ...exhaust gas in the tokamak demonstration reactor (DEMO) consists of more than 90% of unburned fuel gas (D and T) and the remaining part will be He, PEGs and impurities. In DEMO reactor it is foreseen to recover the fuel gas and PEGs. For this purpose in this study a carbon molecular sieve (CMS) membrane and its precursor substrate have been tested at a permeation apparatus built at ENEA Casaccia laboratories. Single gas permeances for H2, He, N2 and Ar were measured with a pressure drop across the membranes between 10 Pa up to 140 kPa at room temperature, 144 °C and 250 °C. We found that H2 to N2 selectivity at 250°C is equal to 47.68 while for H2 to Ar is equal to 35.05.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UILJ, UL, UM, UPCLJ, UPUK, ZAGLJ, ZRSKP
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