Blanket Segments Research Articles

Investigation of MHD flow balancing in electrically conductive manifolds of liquid metal blankets

In the liquid metal (LM) blanket design for magnetic confinement fusion reactors, manifolds including main and branch ducts are commonly used to direct fluids to various locations of the blanket to cool its components. The primary objective of this structural configuration is to achieve flow balancing within each branch duct. However, magnetohydrodynamic effects will cause uneven flow distribution, consequently, uncontrollable overheating phenomena in certain blanket segments may be caused by this flow distribution. To assess its impact, the influences of external magnetic fields, inlet velocities and coupling wall thicknesses on the flow partitioning in conductive manifolds are investigated in this study. The results indicate that the flow distribution among the branch ducts becomes less uniform as the external magnetic field strength increases, while varying the inlet velocity has minimal impact. In addition, increasing the coupling wall thickness leads to more even flow distribution when the rectangular branch ducts are perpendicular with the magnetic field direction. Based on above physical understanding, three controlling methods for the flow partitioning in conductive manifolds are proposed to achieve flow balancing in each branch duct and then validated through numerical simulations. The results indicate that flow balancing can be achieved by adjusting the coupling wall thickness when the rectangular branch ducts are aligned with the magnetic field direction, however, it is challenging when they are perpendicular. Furthermore, the flow balancing can be realized by a symmetrical design of the manifold with circular branch pipes and it is independent of external magnetic fields and inlet velocities. These findings will enhance the physical understanding of flow distribution patterns in conductive manifolds under strong magnetic fields, which is important for the flow control and design of manifolds in LM blankets.

Conceptual design of DEMO breeding blanket in-vessel toroidal transporter

The development of a viable maintenance strategy for the replacement of the Breeding Blanket (BB) segments of DEMO is a key aspect in the way for fusion energy. Previous work concerned the development of a BB vertical transporter to lift, tilt and handle the BB segment inside the remote handling (RH) ports. A potential solution for the replacement of the BB segments could consist in lifting all BB segments from four RH ports, by mean of BB vertical transporter coupled with a BB toroidal in-vessel transporter avoiding the opening of the remaining ports during maintenance. In this way, the activities concerning the removal and re-installation of the auxiliary components (BB feeding pipes, permanent structure for pipes handling, closure plate, port plug, etc.) inside the non-remote handling ports will be avoided, having a strong impact on the reduction of the remote maintenance times and hence on the overall availability of the machine. To assure the feasibility of the proposed strategy a dedicated tool, capable to move toroidally the huge BB segments, shall be developed. The main function of the tool will consist in lifting and toroidal translation of the BB segments that have to be aligned to the RH port, where they will be grabbed and lifted by the BB vertical transporter. The work here presented focuses on the conceptual design of a toroidal transporter to lift and toroidally translate the BB segments inside the Vacuum Vessel. The conceptual design of the BB Toroidal Transporter (BBTT) was developed to handle both the inboard and the outboard segments; a foldable lower foot, equipped with toroidal trucks and suspension system, has been designed to withstand the assumed loads and seismic loads. Preliminary FEM analysis has been carried out to check the structural integrity of the proposed design.

Investigation into optimal feedforward control for remote handling of tokamak breeding blankets

Remote handling of breeding blankets poses an unprecedented challenge in future tokamaks like EU-DEMO, where individual blanket segments can weigh upwards of 80 tonnes and extend beyond 10 m in length. The unparalleled scale of these components, coupled with extremely tight positional tolerances, demands careful consideration of structural flexibility during manoeuvres. This underscores the need for model-based control systems capable of mitigating oscillations and static deflections induced by gravity and disturbances.In this paper, we present recent experiments towards this objective, focussed on mitigating the effects of gravity. We applied model-based, optimal feedforward control to the motion of a planar slender payload, chosen to represent the flexible part of a candidate breeding blanket segment. We combined a predictive path planning algorithm with input shaping techniques to manoeuvre the payload effectively accounting for the influence of gravity, while also reducing post-manoeuvre oscillations. Our results highlight the advantages of model-based control and the limitations associated with the use of feedforward control.

Analysis of implementing a rail-based maintenance system into a HELIAS 5-B device

This paper reports on the analysis of potential rail-based maintenance systems when implemented into a Helical Advanced Stellarator (HELIAS) 5-B device. The main purpose of such a system would be to handle and exchange the internal vessel components, namely the breeding blanket segments, which are expected to be the largest and heaviest components within the vessel. Other rail-based maintenance systems for tokamak devices, particularly the ITER Blanket Remote Handling System (BRHS), were studied, and their suitability when introduced to the differing constraints of a HELIAS device was determined. Rail-based systems envisaged for DEMO and CFETR devices were also studied. An ITER-like handling system was shown to be unsuitable for a HELIAS device of this scale. This is due to significantly higher blanket masses with an in-vessel operating space no larger than that of ITER, but with the additional complication of longer toroidal lengths and non-uniform vessel geometry. Movement of large components on rails like those found in DEMO and CFETR was shown to be more applicable to the HELIAS device. The main obstacle to the implementation of such rails would be the non-uniform geometry which complicates the selection of a rail path that avoids collisions with other in-vessel components, remote handling equipment, or the vessel structure itself.

Progress in the development of the in-vessel transporter and the upper port cask for the remote replacement of the DEMO breeding blanket

The breeding blanket (BB) segments are by far the largest in-vessel components of DEMO. For their remote replacement through the upper vertical ports of the vacuum vessel (VV) recently a new concept has been developed, [1]. The concept minimizes the spread of contamination as all in-vessel operations are carried out from within a cask that is sealed to the VV and located within a sealed room providing a second confinement barrier inside the nuclear building. The removal of the BB segments from the VV is carried out by a BB transporter that is operated on the elevator system of the >20m higher cask. The limited available space makes the compact design solutions that have been developed critical to the overall concept. The BB transporter is designed according to nuclear design codes and for high payloads since the BB segments may weigh up to 180 tons. Due to the eccentric engagement points on the backside of the BB segments and due to seismic accelerations, that need to be considered, too, the BB transporter resists also to bending moments. It can carry out translational as well as tilting movements as required to disengage the BB segments from their supports and to remove them through the upper VV port. The main requirements regarding integration, BB manipulation and structural integrity have been verified. Next development steps need to include further design improvements, integration of in-vessel position survey, definition and control of motion actuations, supply cable routing, the development of rescue and recovery scenarios as well as the validation in relevant test facilities. This article describes the design of the BB lifting tools including several modifications following a set of analyses that were recently performed.

Exploratory study of the EU-DEMO Water-Cooled Lithium Lead breeding blanket behaviour in case of loss of cooling capability

Within the framework of the European Roadmap to the realization of fusion energy, a strong international cooperation is ongoing to develop a Breeding Blanket (BB) system for the EU-DEMO reactor. Although it is still to be decided whether the DEMO in-vessel components should perform any safety function, the pursuing of robust blanket concepts able to handle upset and accidental loading conditions has been always seen as good practice in fusion reactor engineering to enhance the inherent plant safety performances. Amongst the several classes of events that might challenge the BB structural integrity, the large Loss of Coolant Accident is one of the most relevant because it usually leads to a fast loss of cooling capability of the structures. Due to the characteristic of the tokamak assembly, the behaviour of each blanket segment during a sudden loss of cooling capability does not depend only upon distinguishing features of the component itself. In fact, the overall transient can be governed by conditions established in surrounding elements, like adjacent blanket segments and vacuum vessel, as well as by the plasma shutdown strategies adopted to protect the reactor.The scope of the activity herein presented is to make a preliminary assessment of the intrinsic capability of EU-DEMO tokamak architecture to cope with the loss of cooling in the Water-Cooled Lithium Lead (WCLL) BB concept. Evaluation of BB thermal field in short and medium term under simplified, yet conservative, assumptions was carried out for four transient scenarios with the aim of investigating the response of the structure in case of: a) fast or soft plasma shutdown, and b) different blanket cooling schemes. Moreover, the WCLL BB thermo-mechanical response in the most critical time steps has been assessed. The obtained results shall help for future decisions on safety systems/action to be implemented to cope with accidents.

Status of maturation of critical technologies and systems design: Breeding blanket

The scope of the EUFOfusion Work Package Breeding Blanket is to develop a blanket concept for the EU DEMO reactor; this includes the blanket segments inside the Vacuum Vessel and the related Tritium Extraction/Removal Systems. In the Pre-Concept Design (PCD) Phase, two concepts have been selected as candidates; a solid and a liquid breeder blanket cooled with helium and water, respectively. The design of these two blanket systems has been adapted to the DEMO plant design developed in the PCD Phase and performances assessed. A large R&D programme has been implemented with the scope to evaluate different technologies for these blankets; including the development of breeders, tritium extraction and cooling technologies, and the manufacturing of the blanket system. A major milestone in the subsequent Concept Design Phase is the final selection of the blanket concept for DEMO.

Design and integration studies of a diagnostics slim cassette concept for DEMO

The diagnostics slim cassette (DSC) is a concept currently proposed for DEMO to host the microwave reflectometry diagnostic in a dedicated poloidal section, for the feedback control of plasma position and shape, also providing the radial edge density profile at several poloidal angles. This DSC, which is expected to house up to 100 antennas and waveguides, is to be integrated with the breeding blanket (BB) segments. Moreover, it is being designed with remote handling compatibility in mind to facilitate a ‘fast’ exchange by remote maintenance (RM) when needed, fulfiling one of the aims of the work package diagnostic and control. In this approach the pre-assembled (banana-shaped) DSC modules are inserted/removed from the upper port (UP) of the vacuum vessel (VV), at least when the BB segments are replaced. Here the main constraints related to the integration of the DSC with the BB, the UP and the equatorial port are identified and discussed in detail, with a strong focus on the required RM operations inside the VV (in-vessel), providing solutions that can be adapted to present and future blanket and UP designs.

Mechanical support concept of the DEMO breeding blanket

• Description of blanket integration concept. • Stress calculation of breeding blanket segments. • Prediction of Breeding blanket deformation. • Identification of custom-machining concept and prediction of achievable tolerances. • Overview of the various integration issues associated with the design of in-vessel components. The DEMO tokamak architecture is based on large vertical breeding blanket (BB) segments that are accessed from a maintenance hall above the tokamak and are vertically replaced through large upper ports of the vacuum vessel (VV). The feasibility of the BB segments mechanical supports is a prerequisite of this vertical segment architecture. Their design directly impacts on the removal kinematics and the remote handling operations required for release and engagement. The supports must withstand large forces acting on the BB in particular due to electromagnetic (EM) loads. At the same time, they must ensure a sufficiently precise positioning of the BB first wall. Their design also takes into account the significant thermal expansion of the blanket segments that are operated at high temperature avoiding excessive support reaction forces. The BB support concept described in this article does not require fasteners or electrical straps to the VV and therefore much reduces the complexity of the BB remote replacement – a valuable characteristic that would make this concept a milestone in meeting one of the goals defined for the DEMO project: to develop a maintainable fusion power plant design [1] . Each blanket segment is individually supported by the VV without any physical contact to the other blankets or in-vessel components. It relies instead on vertical pre-compression inside the VV due to obstructed thermal expansion and radial pre-compression due to the ferromagnetic force acting on the BB material in the toroidal magnetic field. The verification process did not identify show stoppers. Nonetheless, a further evolution of the concept is required including design improvements to mitigate the high stress levels found in the inboard blankets during plasma disruptions. The fact that no excessively high support reaction forces or large BB deflections were found suggests though that the further development of the concept could be successful.

Conceptual design of the supercritical CO2 cooled lithium lead blanket for CFETR

A conceptual design of the supercritical CO2 cOoled Lithium-Lead (COOL) blanket has been proposed as one advanced blanket candidate for the Chinese Fusion Engineering Testing Reactor (CFETR). At present, the COOL blanket is designed to fulfill the requirement of operating under the fusion power of 1.5 GW of CFETR while realizing the tritium self-sufficiency. Structural designs are carried out for outboard and inboard blanket segments and analyses with regard to neutronics and thermo-hydraulics are conducted to evaluate the blanket performance. Results indicate that a comparatively high Tritium Breeding Ratio (TBR) of 1.183 and a PbLi outlet temperature of 600 – 700°C can be achieved for this blanket concept. Besides, a Power Conversion System (PCS) based on the supercritical CO2 recompressing cycle is preliminarily designed for the blanket and the gross thermal efficiency is estimated to be 39% – 46% when assuming the turbine inlet temperature ranges between 550°C and 650°C.

Integration concept of an Electron Cyclotron System in DEMO

The pre-conceptual layout for an electron cyclotron system (ECS) in DEMO is described. The present DEMO ECS considers only equatorial ports for both plasma heating and neoclassical tearing mode (NTM) control. This differs from ITER, where four launchers in upper oblique ports are dedicated to NTM control and one equatorial EC port for heating and current drive (H&CD) purposes as basic configuration. Rather than upper oblique ports, DEMO has upper vertical ports to allow the vertical removal of the large breeding blanket segments. While ITER is using front steering antennas for NTM control, in DEMO the antennas are recessed behind the breeding blanket and called mid-steering antennas, referred to the radially recessed position to the breeding blanket.In the DEMO pre-conceptual design phase two variants are studied to integrate the ECS in equatorial ports. The first option integrates waveguide bundles at four vertical levels inside EC port plugs with antennas with fixed and movable mid-steering mirrors that are powered by gyrotrons, operating at minimum two different multiples of the fundamental resonance frequency of the microwave output window. Alternatively, the second option integrates fixed antenna launchers connected to frequency step-tunable gyrotrons. The first variant is described in this paper, introducing the design and functional requirements, presenting the equatorial port allocation, the port plug design including its maintenance concept, the basic port cell layout, the transmission line system with diamond windows from the tokamak up to the RF building and the gyrotron sources.The ECS design studies are supported by neutronic and tokamak integration studies, quasi-optical and plasma physics studies, which will be summarized. Physics and technological gaps will be discussed and an outlook to future work will be given.

Integration of service pipes into the lower port for the DEMO Double Null Concept Design Study

A study into the viability of a double null concept for DEMO has been undertaken as part of the work to address Key Design Integration Issues within the DEMO powerplant. The objective of this study was to evaluate the feasibility of a double null architecture featuring split blanket segments to enable a wider range of maintenance options.One of the main challenges found was the integration of the service pipes to provide connections to the cooling and tritium breeding systems for the divertor and breeding blanket. By understanding how these pipes might be integrated, the constraints on key details such as the length of the lower port could be understood. This allowed for progression of other areas of the design and investigations into the remote handling of these pipes during maintenance periods.The key requirements identified for the service pipe routing included determining the removal sequence for the blanket and divertor segments during maintenance; access for the cutting and welding tool at the pipe/blanket interface; pipes grouped as modules per component; and optimisation of port space for the vacuum pumps.Two proposals were considered to investigate their impact on the insertion location for the cutting and welding tool: pipes with large radii, and pipes with smaller radii. The former was found to be preferable due to the remote maintenance advantages, versus the limited space gains offered by the latter. This paper will describe the development of these requirements, both proposals, and discuss the challenges uncovered during the study in more detail.

The development of a Sustained High Power Density (SHPD) facility

A Sustained High Power Density (SHPD) facility is being planned within the U.S. as a non DT device which can be a bridge to a Compact Fusion Pilot Plant (CFPP) DT operated facility with the future capacity to produce a levelised cost of electricity at a rate competitive with current power generating systems. To accomplish this a review of ITER cost data and recent next step physics/engineering studies that promoted technology development, physics enhancements and innovated configuration improvements that increase operating availability have been injected into a proposed PPPL SHPD design. Much of the physics and engineering design basis for this effort centers on a series of ST studies carried out in recent years attempting to maximize the low aspect ratio (AR) ST performance within realistic engineering constraints of this compact fusion device. One issue found in the low AR design is the lack of volt-seconds generated by a small central solenoid that resides in the available space left after sizing the TF structure to support the high current density/HTS TF coil. Although not required in a larger pilot plant device, a solenoid wrapped around the plasma side of the SC TF coil was considered for the small low aspect ratio SHPD device. Adding a solenoid to the plasma side of the TF imposes a major change to the device configuration – a change which has been found to have cost and assembly issues when compared with an OH solenoid located at the machine center. The inclusion of a liquid metal first wall divertor system, the integration of local pairs of outboard DCLL blanket segments and a summary of physics performance conditions will be presented.

Interface design of the blanket system for CFETR

The China Fusion Engineering Test Reactor (CFETR) aims to bridge the gaps between fusion experimental reactor ITER and demonstration fusion power reactor. Blanket, one of the key components of the CFETR, is of great importance to achieve the functions of tritium breeding, heat extraction and radiation shielding. As for the next device for fusion energy research in China, two blanket candidate options have been developing in parallel to decrease the risk of both physical and technical issues. Apart from the research and design work on the breeding blanket module itself, the interfaces definition and integration of the blanket system is vital, particularly due to the system interdependencies. The interface design of the blanket system is a challenge due to several factors, such as the thermal expansion, the electromagnetic loads, the feasibility of remote maintenance. In this paper, a preliminary interface definition and integration design of the CFETR blanket system has been presented, based on the design with R = 7.2 m and a = 2.2 m. The purpose is to complete the interface definition and integrated design of the blanket and other components or systems. The vacuum vessel has 16 sectors, and each sector has two inboard and three outboard blanket segments. The blanket segment will be installed and removed by remote maintenance tools through the upper ports. The work addresses the interfaces between the blanket and other system, such as diagnostic system and heating system. Preliminary assessment of the impact on tritium breeding ratio (TBR) for the candidate water cooled ceramic breeder blanket integration scheme has been presented highlighting the influence of the auxiliary interfaces system to TBR.

Preliminary design of the top cap of DEMO Water-Cooled Lithium Lead breeding blanket segments

Within the framework of EUROfusion R&D activity, a research campaign has been carried out at the University of Palermo, in close cooperation with ENEA labs, in order to preliminary design the top cap foreseen for the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket segments. Due to the high heat and pressure loads acting on such component, its design results particularly demanding and a specific multi-physics approach is needed, covering several aspects from thermal-hydraulics to structural assessments. Preliminary detailed CAD model of the cap integrated into the upper region of the WCLL breeding blanket outboard central segment has been set-up, equipped with proper cooling circuits as well as manifold and attachment systems according to the design of the equatorial elementary cell. A detailed numerical model has been set-up with the aim of simulating the thermo-mechanical behaviour of the above-mentioned system. Finally, the thermo-mechanical response of the upper breeding blanket region has been evaluated in terms of stress and temperature distributions, verifying that the structural material maximum temperature stays below its limit value and that structural integrity is ensured by means of the fulfilment of design rules reported in RCC-MRx structural design code.

Key design integration issues addressed in the EU DEMO pre-concept design phase

The EU fusion roadmap defines the development of a DEMO plant that achieves a long operation time and demonstrates tritium self-sufficiency and net electricity output. The pre-conceptual design phase will be concluded with a gate review in 2020. Eight key design integration issues have been identified as critical, either because the corresponding solution found in ITER is not suitable in DEMO or because the issues are DEMO-specific and not present in ITER. All of these will affect in their resolution the design and possibly the technology of several tokamak and plant systems or even the DEMO architecture: (i) Feasibility of wall protection limiters during plasma transients, (ii) integrated design of breeding blanket and ancillary systems, (iii) power exhaust taking advantage of advanced divertor configurations, (iv) tokamak architecture based on vertical blanket segments, (v) direct or indirect power conversion concept, (vi) configuration of plant systems in the tokamak building, (vii) feasibility of hydrogen separation in the torus vacuum pump and direct internal fuel recycling, and (viii) development of a reliable plasma scenario.

Status of the DEMO blanket attachment system and remaining challenges

The DEMO blanket attachment concept is challenging due to several factors: the harsh radiation environment, the thermal expansion, the electro-magnetic loads, the remote maintenance feasibility, and the accurate control of the alignment of the breeding blanket first wall during operation. There are two inboard and three outboard blanket segments per vacuum vessel sector to be installed and extracted by remotely controlled tools through a single upper vertical port. The design of the fixations of the blanket segments to the VV complies with the strategy to avoid the need for front side access for engagement and release. The attachment system has been designed for the numerous critical load cases, including normal operation, dwell between pulses, plasma disruptions, fast discharge of the magnet coils, and accidental conditions such as loss of blanket coolant. At the same time the attachments must guarantee the stresses in the blanket segments not to exceed limits.This paper introduces the attachment concept and describes the finite element model that has been built to assess the blanket attachment system. The model represents one sector of the DEMO machine. The results focus on the reaction forces transmitted at individual attachment locations to define these interfaces and guide the design of the individual supports.

Nuclear analysis of the Single Module Segment WCLL DEMO

The latest design of the Water Cooled Lithium Lead blanket for the European DEMO reactor is based on a Single Module Segment (SMS) concept for the blanket segments: the removal of the gaps between the modules (present in the former Multi Module structure design), with the consequent reduction of the neutron streaming, improves the tritium production and the shielding capabilities. A nuclear analysis has been carried out with the Monte Carlo N-Particle transport code version 5 (MCNP5) and the Joint Evaluated Fission and Fusion nuclear data libraries version 3.2 (JEFF 3.2). A detailed three-dimensional MCNP DEMO model with the Single Module layout has been generated. Three-dimensional neutron and gamma transport simulations have been carried out in order to assess the tritium production, nuclear power deposition in each subcomponent of the blanket and nuclear heating density, neutron flux, neutron damage (dpa) and Helium production, relevant for the design of the system. The results confirm the fulfilment of tritium self-sufficiency and shielding requirements of DEMO. Furthermore, a parametric analysis has been carried out by varying the thickness of the first wall tungsten layer to evaluate how the blanket performances are influenced by the tungsten layer thickness.

EU-DEMO Breeding Blanket temperature evaluation before remote maintenance operation

In DEMO (the European proposal of demonstration fusion reactor) the Breeding Blanket (BB) segments shall be periodically replaced, at end of life or after failure. Any replacement of a BB segment shall be performed by Remote Handling Equipment (RHE) through the vertical upper port: any active action is executed by RHE that clamps a BB segment in the upper portion using a BB transporter. A driving requirement of DEMO commercial technology for RHE is the maximum operating temperature at the BB-RHE interface during BB replacement: the objective of this study is properly to assess this limit, being now the allowable value at 100 °C. This study deals with the Helium Cooled Pebble Bed (HCPB) concept.The initial configuration, with all sector segments (three in the outboard region and two in the inboard one) in their nominal position, has been considered; then the Finite Element model predicts if a natural convection cooling air flux inside vacuum vessel is adequate or if it is necessary a forced convection cooling system. The decay heat produced by the main blanket components (the BB modules caps and lateral walls, the backwalls, the supporting structures, the backplate manifolds, and so on) have been evaluated beforehand: the values applied as body loads in the thermal analyses have been selected within the set related to one month after shut-down. With the assumed geometrical and physical boundary conditions set for the natural and forced convection, the analyses indicate that a forced convection should be necessary to ensure compliance with the present RHE requirements. The obtained results, the analysis assumption and future analysis plan have been briefly discussed.

Initial configuration studies of the upper vertical port of the European DEMO

In the current pre-concept phase of the European DEMO, integration studies of the systems in the Upper Port area are being carried out. In DEMO, the Upper Port of the Vacuum Vessel is extraordinarily large to allow for the vertical extraction of the Breeding Blanket segments. This requires a number of components inside and outside the port to be integrated with tight space constraints: The Upper Port structure and its annexes, the adjacent Toroidal and Poloidal Field Coils, the Thermal Shields, the piping connection to the Vacuum Vessel Pressure Suppression System, the Shield Plug and its inserts, the feeding pipework of the in-vessel components and part of the Breeding Blanket supporting structures.Apart from functional aspects, the design of these components is driven by considerations of structural integrity, maintainability and irradiation shielding, which are mutually competing in many areas. Several studies were conducted on the design of the Upper Port and the required configuration of the components within. The present article describes the development approach, the studied options and the respective results, the identified issues as well as the proposed engineering solutions, in particular with respect to the mechanical design of the Upper Port and the integrated Shield Plug.