Wall Blanket Research Articles

Induction brazing of DEMO's large bore cooling pipes

The DEMO first wall blanket components will be exposed to high neutron fluence and high levels of thermal energy, the latter being removed by a coolant fluid transmitted out of the reactor via service piping. Due to the thermal and nuclear loading on the first wall, the components will need to be periodically removed and replaced. Installation of new blankets requires the joining of their service pipes by remote handling.Brazing is one of the alternatives that are considered for joining the service pipes of the DEMO breeding blankets. This technique has great potential and has several major advantages over welding.The Fusion Technology Department of HUN-REN Centre for Energy Research (CER) has been working on the development of a viable brazing concept of thick, large bore pipes inside the ports of DEMO. Induction has been chosen as the heat source for the brazing operation, which is a well-studied technology and has many uses in the industry already. However, brazing in this size is a completely new horizon for induction heating so CER's work includes a significant amount of development. Firstly, the coil is naturally a bespoke component, which has been optimized by electromagnetic analysis. Secondly, a test rig has been designed and manufactured in-house by CER, to test the performance of the induction heater in a real scenario.This paper outlines the development strategy of the induction brazing of DEMO's large bore pipes.

GAMBL – A dual-cooled fusion blanket using SiC-based structures

We describe a novel fusion reactor blanket concept called GAMBL – GA Modular BLanket. This design concept exploits the advantages of SiC-based structures while minimizing their limitations. Material microstructures are tailored for different functions throughout the first wall and blanket, including the use of graded W/SiC composites to enhance heat transfer capabilities and resist erosion at the plasma-facing surface. The design contains a decoupled first wall and breeding blanket, which we show can provide adequate tritium breeding. By decoupling the breeding part of the blanket, we can operate at very low PbLi breeder pressure and support gravity loads in simple, radiatively-cooled structural beams that do not contain coolant. First wall performance capabilities are enhanced by eliminating constraints imposed by the deep blanket. We have performed initial fluid, thermal, mechanical and neutronics analysis of this concept to show its capabilities to meet the design requirements. R&D needs are mostly related to materials development. Except for unique manufacturing needs, no new facilities are required beyond existing or planned facilities for the US base program on steel-based DCLL blankets.

Modeling Magnetic Material Using Vector Potential-Based Elements in ANSYS and Comparing with Edge-Based Elements

More and more, soft or hard magnetic materials have been used in the magnetic confinement fusion machine, like reduced-activation ferritic-martensitic steel, which may be used in the first wall and breeding blanket structural material of ITER. Twisted coils in the stellarator may be substituted by permanent magnets. ANSYS electromagnetic simulation has two methods: nodal-based formulation or edge-based formulation. Nodal-based formulation has been widely used in past years. This paper will present a modification to the nodal-based element SOLID97 in ANSYS to form the correct normal boundary condition to model ferromagnetic materials and compare the results, e.g., field, eddy current density, and Lorentz force, to edge-based methods using SOLID236.

Development of anti-permeation and corrosion barrier coatings for the WCLL breeding blanket of the European DEMO

Tritium permeation from breeder material to the Water Coolant System (WCS) in Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) is one of the technological issues to be solved in the design of the European DEMO. Since the tritium extraction from the Water Coolant System is more challenging and expensive than the extraction from the eutectic alloy PbLi, it is mandatory to use of a protective coating on the blanket wall to minimize the permeation rate. Moreover, a protective coating can prevent the corrosion of EUROFER steel by the action of PbLi.alumina-based coatings are considered as reference for barriers thanks to their good chemical compatibility with the PbLi alloy and their capability to reduce permeation. Three coating technologies were selected in the frame of the EUROfusion project: electrochemical ECX (chemical deposition) process, Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) coating. The coatings were developed and optimized in order to satisfy the design requirements of good mechanical compatibility with steels, strong adhesion, corrosion compatibility in PbLi at relevant BB design conditions and a Permeation Reduction Factor at least of 200 under neutron irradiation. The present paper aims to describe the status of the technologies and the main results obtained.The final objectives of the R&D activities are to demonstrate the applicability of the coating to WCLL BB and therefore the scale-up of the technologies from laboratory scale to the BB scale.

Effect of nano-coating on corrosion behaviors of DCLL blanket channel

The corrosive behaviors of liquid metal fluid flow on the channel wall of Dual Coolant Liquid (DCLL) blanket of fusion reactor were simulated by imposing the discrete phase model (DPM) under different magnetic field strength, inlet velocities and neutron heat sources while considering blanket channel walls with and without non-wetting nanomaterial coating conditions. Results showed that regardless of the external environment of the blanket (except in the absence of magnetic field), the corrosion of the inner walls of the blanket channel declined significantly after employing nano-coatings in comparison to that of uncoated wall. For the uncoated condition, channel wall corrosion rate increased exponentially with time, various parameters have significant influence on channel corrosion, and the effect of magnetic field intensity is highest among them. In contrast, the corrosion rate of the channel walls with nanomaterial coating increased linearly with time, and unlike uncoated wall there is little difference on the corrosion rate of the channel walls for different parameters. Hence, these minute alterations in the corrosion of the channel walls with varying parameters indicate that the nanomaterial coating layer has a substantial protective effect on the blanket wall structure.

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

Effects of silicone rubber and aerogel blanket-walled tubes on H2/Air gaseous detonation

The experimental study of hydrogen-air detonation was performed in a 6.3-m-long circular cross-section tube, the rear section of which was walled by silicone or aerogel material. Mixtures of 25% or 29.6% hydrogen with air by volume with an initial pressure of 0.1 MPa and temperature of 293 K were used in the experiments. A short Shchelkin spiral was fixed at the tube head to accelerate the deflagration-to-detonation transition. The flame and pressure wave propagation velocities were measured by ion probes and pressure transducers, respectively. The detonation regimes and velocities of flame propagation in walled tubes of different materials were compared. It was found that it is difficult for a silicone rubber-walled tube to reduce the intensity of the detonation wave. In this case, for the stoichiometric mixture, a reduction in the propagation velocity by 5% and the pressure peak by 12% was recorded in order to compare those parameters in a stainless smooth tube. An aerogel blanket wall has a strong ability to reduce the intensity of the detonation wave with the peak pressure decreased by 35% for the stoichiometric hydrogen-air mixture. The attenuation of the steady detonation wave can occur in the aerogel blanket-walled tube under the present experimental conditions. This is because of the effective suppression of transverse waves by interaction with the aerogel wall; then, the generation of stable detonation cells is no longer possible, and thus, the shock wave and flame front are decoupled, and detonation is suppressed.

The calculation of neutron flux using Monte Carlo method

In this study, a hybrid reactor system was designed by using 99–95% Li20 Sn80 + 1-5% RG-Pu, 99–95% Li20 Sn80 + 1-5% RG-PuF4 , and 99–95% Li20 Sn80 + 1-5% RG-PuO2 fluids, ENDF/B-VII.0 evaluated nuclear data library and 9Cr2WVTa structural material. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion–fission hybrid reactor system. The neutron flux was calculated according to the mixture components, radial, energy spectrum in the designed hybrid reactor system for the selected fluids, library and structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

Effect of thick blanket modules on neoclassical tearing mode locking in ITER

The rotation of m/n = 2/1 tearing modes can be slowed and stopped (i.e. locked) by eddy currents induced in resistive walls in conjunction with residual error fields that provide a final ‘notch’ point. This is a particular issue in ITER with large inertia and low applied torque (m and n are poloidal and toroidal mode numbers respectively). Previous estimates of tolerable 2/1 island widths in ITER found that the ITER electron cyclotron current drive (ECCD) system could catch and subdue such islands before they persisted long enough and grew large enough to lock. These estimates were based on a forecast of initial island rotation using the n = 1 resistive penetration time of the inner vacuum vessel wall and benchmarked to DIII-D high-rotation plasmas, However, rotating tearing modes in ITER will also induce eddy currents in the blanket as the effective first wall that can shield the inner vessel. The closer fitting blanket wall has a much shorter time constant and should allow several times smaller islands to lock several times faster in ITER than previously considered; this challenges the ECCD stabilization. Recent DIII-D ITER baseline scenario (IBS) plasmas with low rotation through small applied torque allow better modeling and scaling to ITER with the blanket as the first resistive wall.

Phase separation and ω transformation in binary V-Ti and ternary V-Ti-Cr alloys

V-Ti-Cr alloys have emerged as the potential candidate for many advanced engineering applications e.g. first wall blanket for fusion reactors, alternative energy technologies and hydrogen storage. Phase transformation behaviour of these alloys is not completely understood and the binary V-Ti phase diagram is still debated. The system becomes even more complicated with the addition of Cr as a ternary addition. Thermodynamic calculations indicate that within certain composition range in the V-Ti system, a prominent miscibility gap exists and the propensity for phase separation decreases with the addition of Cr in the binary system. Alloys with nominal composition of V-50Ti and V-40Ti-20Cr have been prepared through vacuum arc melting and detailed microstructural and microchemical studies have been carried out. Diffraction contrast and phase contrast microscopy studies confirm the occurrence of phase separation in both the alloys. The lamellar and interconnected nanostructured domains with coherent interfaces as seen in the phase contrast mode are indicative of possible phase separation in the alloys. It has further been confirmed by High-angle annular dark field (HAADF) studies associated with STEM-XEDS and STEM-EELS. Further, microscopic studies on the as melted and the heat treated alloys confirm that the phase separation propagates through a spinodal decomposition mechanism in this alloy and is not limited by nucleation and growth kinetics. Another interesting aspect of this study is the formation of nanostructured ω phase in the phase separated regions. Phase contrast microscopy studies in selected orientations confirm the presence of distorted hexagonal zones of ω phase in the phase separated domains of the alloys.

Assessment of the measurement performance of the in-vessel system of gap 6 of the ITER plasma position reflectometer using a finite-difference time-domain Maxwell full-wave code.

We conducted a first assessment of the measurement performance of the in-vessel components at gap 6 of the ITER plasma position reflectometry with the aid of a synthetic Ordinary Mode (O-mode) broadband frequency-modulated continuous-wave reflectometer implemented with REFMUL, a 2D finite-difference time-domain full-wave Maxwell code. These simulations take into account the system location within the vacuum vessel as well as its access to the plasma. The plasma case considered is a baseline scenario from Fusion for Energy. We concluded that for the analyzed scenario, (i) the plasma curvature and non-equatorial position of the antenna have neglectable impact on the measurements; (ii) the cavity-like space surrounding the antenna can cause deflection and splitting of the probing beam; and (iii) multi-reflections on the blanket wall cause a substantial error preventing the system from operating within the required error margin.

Effect of design geometry of the demo first wall on the plasma heat load

In this work we analyse the effect of W armour surface shaping on the heat load on the W/EUROFER DEMO sandwich type first wall blanket module with the water coolant. The armour wetted area is varied by changing the inclination and height of the «roof» type armor surface. The deleterious effect of leading edge at the tiles corner caused by misalignment is replaced in current design by rounded corners. Analysis has been carried out by means of the MEMOS code to assess the influence of the thickness of the layers and effect of the magnetic field inclination. Calculations show the evolution of the maximum temperatures in the tungsten, EUROFER, Cu allow and the stainless-steel water tube for different level of surface inclination (chamfering) and in the case of rounded corners used in the current design. It is shown that the blanket module materials remain within a proper temperature range only at shallow incident angle if the width of EUROFER is reduced at list twice compare with the reference case.

Effect of heat loads on the plasma facing components of demo

In this paper we analyse a thermo-hydraulic performance of the first wall blanket module during the stationary DEMO operation with the edge localized mode (ELM). Heat loads are estimated based on scaling arguments and predictions from the peeling-ballooning ELM model. Effect of parallel heat fluxes intersecting with the first wall panels and avoidance of overheating by inclination of the panels are considered. The material temperatures of the W/EUROFER sandwich type module with water cooling stainless stell tube and Cu alloy compliance embedded into EUROFER is calculated by using the MEMOS code. The calculations were carried out indicating the required geometric parameters as well as the cooling conditions which allow keeping materials temperatures within allowable engineering limits. Effect of inclination of the first wall plates to avoid the misalignment problems is considered.

The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system

AbstractIn the present investigation, a hybrid reactor system was designed by using 100 % Flibe, 90 % Flibe-10 % ThF4, 90 % Flibe-10 % UF4 fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion–fission hybrid reactor system. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

Thermo-fluid dynamics and corrosion analysis of a self cooled lead lithium blanket for the HiPER reactor

The HiPER reactor is the HiPER project phase devoted to power production. To reach a preliminary reactor design, tritium breeding schemes need to be adapted to the HiPER project technologies selection: direct drive ignition, 150 Hz of power released through fusion reactions, and the dry first wall scheme. In this paper we address the main challenge of the HiPER EUROFER-based self cooled lead lithium blanket, which is related to the corrosive behavior of Pb–15.7Li in contact with EUROFER. We evaluate the cooling and corrosion behavior of the so-called separated first wall blanket (SFWB) configuration by performing thermo-fluid dynamics simulations using a large eddy simulation approach. Despite the expected improvement over the integrated first wall blanket, we still find an unsatisfactory cooling performance, expressed as a low outlet Pb–15.7Li temperature plus too high corrosion rates derived from local Pb–15.7Li high temperature and velocity, which can mainly be attributed to the geometry of the channels. Nevertheless, the analysis allowed us to devise future modifications of the SFWB to overcome the limitations found with the present design.

Mary Walker Phillips: The Art of Knitting

Mary Walker Phillips (1923–2007) was among a postwar generation of artists trained in the use of craft materials and techniques, including textiles, ceramics, glass, wood, and metals, at a handful of progressive art schools forged on European models. At Cranbrook Academy of Art, Phillips and her peers at midcentury bridged industrial design and studio craft to create beautiful, affordable, well-designed products and works of art that defined American interiors and contemporary lifestyles. Drawing upon research for my MA thesis, Mary Walker Phillips: “Creative Knitting” and the Cranbook Experience, I document Mary Walker Phillips’ foray into knitting as creative medium. Starting in 1964 with the 13th Milan Triennale, Phillips’ abstract knitted wall hangings, cushions, blankets, and lampshades were featured in notable exhibitions of art, craft, and interior design. A resident of New York and California, Phillips’ ground-breaking work in knitting in the 1960s reflects the visual legacy of Expressionism, and of bi-coastal, urban, and international influences on American modernism and studio craft.

ITER Diagnostic First Wall

The ITER Diagnostic Division is responsible for designing and procuring the First Wall Blankets that are mounted on the vacuum vessel port plugs at both the upper and equatorial levels. This paper will discuss the effects of the diagnostic aperture shape and configuration on the coolant circuit design. The Diagnostic First Wall (DFW) design is driven in large part by the need to conform the coolant arrangement to a wide variety of diagnostic apertures combined with the more severe heating conditions at the surface facing the plasma, the First Wall (FW). At the FW, a radiant heat flux of 35W/cm2 combines with approximate peak volumetric heating rates of 8W/cm3 (equatorial ports) and 5W/cm3 (upper ports). Here at the FW, a fast thermal response is desirable and leads to a thin element between the heat flux and coolant. This requirement conflicts with the desire to have a thicker FW element to accommodate surface erosion and other off-normal plasma events.

Thermal hydraulic feasibility analysis of the IBED PHTS for ITER

One of the main challenges of the ITER fusion reactor is to effectively remove large amount of heat deposited to the surface of the plasma facing components. The tokamak cooling water system (TCWS) will accomplish the objective of removing about 1GW of peak heat load from in-vessel components while maintaining pressures and temperatures of the coolant within acceptable and safe limits during different operational scenarios. A study of feasibility has been launched for the IBED PHTS (Integrated Blanket, Edge localized mode coils (ELMs) and Divertor Primary Heat Transfer System; it consists of five independent cooling trains (four operational and one in stand-by), one steam pressurizer, supply and return headers, ring manifolds and connections to the all in-vessel components (i.e. First Wall Blanket, Divertor, ELM, Diagnostics and other Ports clients).The dynamic behaviour of the IBED PHTS has been investigated by means of RELAP5® code to simulate the response of the system during plasma pulse and baking operations. Due to the plasma heat deposition on the surfaces of the in-vessel components and subsequent increase in hot leg temperature, a large amount of water volume is transferred from the hot legs of the circuit to the surge-line of the pressurizer during each burn cycle. This causes rapid increase of pressure and temperature of the system and the following actions are proposed to counteract these variations: spray injection in the upper dome of the pressurizer from the Chemical and Volume Control System (CVCS) to reduce the pressure and active control of flow rates through heat exchangers and their bypass loops to regulate the heat transfer from the primary system to the environment via secondary and tertiary loops.This paper focuses on the prediction of the thermal hydraulic behaviour of the IBED PHTS during plasma pulses and baking scenarios, describing the various activity of the analysis, the geometrical assessment of the circuit and the modelling with RELAP5® code. The results have been compared with design and operational requirement. Possible strategies to enhance the system performances have been formulated.

Applicability of tungsten/EUROFER blanket module for the DEMO first wall

In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14MW/m2. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface.

Studies of a self-cooled lead lithium blanket for HiPER reactor

Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall – Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates.