DEMO Breeding Blanket Research Articles

Experimental insights on iron-based alloys corrosion in water cooled loops

This paper presents experimental findings on the behavior of iron-based alloys in environmental conditions typical of nuclear fusion technology, specifically focusing on material degradation, which is a critical aspect for the water cooling system of EU DEMO breeding blankets. The experimental campaign investigates potassium hydroxide’s role as an alkalizing agent, testing various concentrations to assess its impact on corrosion resistance. Additionally, it examines how oxygen levels affect localized corrosion development, which is crucial for mitigating corrosion risks in fusion applications. Seven 1000-hour tests were conducted to determine optimal conditions for corrosion reduction. Findings include identifying an oxygen concentration threshold to prevent piping cracking on EUROFER97 specimens.

Insights from applications of the RSTM tool for coupled CFD-activation fluid simulation

Accurate prediction of the activation of fluids flowing under irradiation is important for the timely development of fusion technology. One of the major current issues is the cooling water of ITER, which becomes activated by plasma neutrons during nuclear operations, thus raising a number of radiological implications such as radiation effects on sensitive equipment, compliance with radiological protection zoning, and compliance with radioactivity limits in regulations of pressure equipment and effluents. Further significant applications are the activation of other service fluids in ITER and of the LiPb in both ITER and DEMO breeding blanket modules.To avoid systematic uncertainties, simulation of the activation of fluids flowing in arbitrarily complex 3D geometry, flow regimes and neutron fields requires full coupling of activation with fluid-dynamics physics models. No such tools were available until recently: the Radio-Species Transport Model (RSTM), based on the well-established ANSYS Fluent® UDS methodology, is one conceived and developed at F4E. Here we review the methodology and capabilities of the RSTM, as well as earlier development, validation, benchmarking and application activities. We then report currently ongoing further applications and benchmarking being performed in collaboration with specific tasks of the EUROFusion programme Preparation for ITER Operation. Computations, results and comparison with other methodologies for several cases of interest are presented and discussed.

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.

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.

RELAP5-based thermal-hydraulic assessment of the STEAM facility for DEMO WCLL balance of plant analysis

DEMO power station aims to demonstrate the generation of hundred MWs of electrical power from fusion reactions, then transmitted from the Tokamak reactor to the grid through the Balance of Plant (BoP). The design approach for the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) leverages nuclear industry expertise but faces challenges due to DEMO pulsed operation and low-load periods. To assess the feasibility of these components, ENEA Experimental Engineering Division at Brasimone R.C. is designing STEAM, a facility investigating water technologies applied to the DEMO BB and BoP systems and components. STEAM is mainly composed of a primary system reproducing the DEMO WCLL BB PHTS thermodynamic conditions (15.5 MPa, 328–295 °C) and a secondary two-phase (liquid/steam) loop reproducing the DEMO power conversion system conditions (6.4 MPa, 238–300 °C). Experimental validation will reproduce steady-state and transient operation under DEMO-relevant conditions, including dedicated tests on the DEMO once through steam generator mock-up.STEAM objectives and description are presented in this paper, together with the RELAP5/Mod3.3 nodalization of the facility. The latter is used to, thermal-hydraulically characterize the facility behaviour. The outcomes of the steady-state qualification supported the optimization of the system layout appraising the performances of key components under the specified operating conditions.

Assessment of the relevancy of ENEA Water Loop facility with respect to ITER WCLL TBS Water Cooling System by considering their thermal-hydraulic performances

The Water-Cooled Lead-Lithium (WCLL) is one of the two candidate concepts for the Breeding Blanket (BB) of DEMO. A Test Blanket Module (TBM) together with its Water Cooling System (WCS) is going to be installed and tested in the ITER reactor. The WCS acts as primary cooling circuit of the TBM module, and it is designed to reproduce the water thermodynamic conditions expected at the DEMO BB inlet.During last years, ENEA and the DIAEE of Sapienza University of Rome have carried out the conceptualization of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone. The W-HYDRA platform is composed by three individual facilities called: Water Loop, Steam, and LIFUS5/Mod4. Water Loop replicates the salient thermal-hydraulic features of the ITER WCLL WCS, and it is equipped with a test section placed inside a Vacuum Vessel (VV) to investigate mock-ups of the whole TBM or its individual parts.This paper assesses the relevancy of the WL facility with respect to ITER WCLL TBM System comparing their thermal-hydraulic performances during selected operational and accidental conditions. Two RELAP5/Mod3.3 models were developed, and the outcomes of the simulations showed a good agreement, thus demonstrating the effectiveness of WL facility to support the ITER TBM program.

Applicability study of mechanical multi-pipe connections for DEMO breeding blanket maintenance concept

Maintenance of DEMO breeding blanket includes the removal and replacement of plasma facing components, for example the breeding blankets (BB). For access, multiple coolant pipes need to be removed. As an option to reduce downtime and increase maintenance speed, a so-called mechanical multi-pipe connection (MPC) concept is developed to allow removal of multiple pipes at the same time using remotely operated mechanical connections. MPC is one promising candidate to allow fast remotely controlled opening and closing of multiple pipe connections by using flanges and bolts.Some major challenges for the application of MPC are highly constrained space on top of the BBs, accessibility for appropriate tools and homogenous distribution of sealing forces on the gaskets. Numerous studies on arrangement and force application methods were performed to identify suitable concepts – of which a compact flange design using direct bolted connection is favourable. Finite Element Analyses on the shape and geometry optimization of the MPC led to analytically verified solutions. Furthermore, material selection and behaviour under radiation, vacuum, and thermal cycling at high temperature conditions requires advancement of available material treatment solutions in terms of things such as coated surfaces of threads and contact areas on flanges to avoid diffusion bonding.

Applicability to DEMO breeding blankets of neutron measurements techniques from fission reactors

Neutron measurements are routinely performed in fission reactors. A lot of experience has been gained at CEA in designing, prototyping and testing neutron sensors and associated acquisition systems in several nuclear facilities with a wide range of demanding conditions. In this paper, the applicability of this experience to neutron diagnostics in the breeding blankets of DEMO is addressed, as a complement to the already planned diagnostic strategy.The breeding blankets are a key component of a fusion reactor such as DEMO. The expected performances and operational constraints are demanding, nevertheless some of those, as the neutron flux range and high temperature, are close to what is encountered in sodium-cooled fast neutron reactors.The strengths and drawbacks of neutron sensors are presented with respect to the performances and constraints. Self-powered neutron detectors and fission chambers benefit from a vast return of experiment in fission reactors and received recent developments in the framework of studies for sodium-cooled fast neutron reactors. The innovative optical fission chamber is discussed for its electromagnetic immunity and high dynamic range demonstrated up to seven decades. Dosimetry with activation foils is also considered for its ability to provide an absolute calibration and spectral unfolding.

Overview of progress on water cooled ceramic breeder blanket in Japan

National institutes for Quantum Science and Technology (QST) are the implementing body of ITER project in Japan and takes the lead in fusion DEMO reactor development in Japan. The primary candidate of the breeder blanket for JA DEMO is Water Cooled Ceramic Breeder (WCCB) blanket. ITER-Test Blanket Module (TBM) program is an opportunity for the demonstration of DEMO breeding blanket concept under actual fusion environment. QST has led the development of the WCCB TBM Concept in the program since the signing of the TBM Arrangement in 2014. The preliminary design of test blanket system (TBS) is on-going. Recently, the status of the preliminary design of TBS was assessed between QST and the ITER Organization (IO). So, the status of the latest TBS design and the DEMO relevancy is presented.

STEAM Experimental Facility: A Step Forward for the Development of the EU DEMO BoP Water Coolant Technology

Within the EUROfusion roadmap for the technological development of the European-DEMOnstration (EU-DEMO) reactor, a key point has been identified in the discontinuous operation (pulse-dwell-pulse) of the machine. Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) adopt technology and components commonly used in nuclear fission power plants, whose performances could be negatively affected by the above mentioned pulsation, as well as by low-load operation in the dwell phase. This makes mandatory a full assessment of the functional feasibility of such components through accurate design and validation. For this purpose, ENEA Experimental Engineering Division at Brasimone R.C. aims at realizing STEAM, a water operated facility forming part of the multipurpose experimental infrastructure Water cooled lithium lead -thermal-HYDRAulic (W-HYDRA), conceived to investigate the water technologies applied to the DEMO BB and Balance of Plant systems and components. The experimental validation has the two main objectives of reproducing the DEMO operational phases by means of steady-state and transient tests, as well as performing dedicated tests on the steam generator aiming at demonstrating its ability to perform as intended during the power phases of the machine. STEAM is mainly composed of primary and secondary water systems reproducing the thermodynamic conditions of the DEMO WCLL BB PHTS and power conversion system, respectively. The significance of the STEAM facility resides in its capacity to amass experimental data relevant for the advancement of fusion-related technologies. This capability is attributable to the comprehensive array of instruments with which the facility will be equipped and whose strategic location is described in this work. The operational phases of the STEAM facility at different power levels are presented, according to the requirements of the experiments. Furthermore, a preliminary analysis for the definition of the control strategy for the OTSG mock-up was performed. In particular, two different control strategies were identified and tested, both keeping the primary mass flow constant and regulating the feedwater mass flow to follow a temperature set-point in the primary loop. The obtained numerical results yielded preliminary feedback on the regulation capability of the DEMO steam generator mock-up during pulsed operation, showing that no relevant overtemperature jeopardized the facility integrity, thanks to the high system responsivity to rapid load variations.

Testing needs for the development and qualification of a breeding blanket for DEMO

The purpose of this paper is to critically review the testing and qualification strategy for the DEMO breeding blanket in Europe, identifying the potential risks and weaknesses and recommending, where necessary, the changes required to strengthen or accelerate the programme. Based on the information presented in this paper, a risk mitigation strategy is required to reduce the technological, licencing, and regulatory risks associated with the performance and reliability of the breeding blanket technologies for DEMO. An attractive mitigation option, which was originally proposed more than 30 years ago but not implemented, is to perform testing and qualification of breeding blanket technologies and design concepts in a dedicated nuclear plasma device that serves as a 14 MeV volumetric neutron source.

Structural assessment of the gripper interlock of the DEMO breeding blanket transporter

The maintenance of the DEMO Breeding Blanket (BB) remotely is a crucial aspect in development of the DEMO power plant. It is a challenge due to the huge mass of the BB segment of about 180 tons. A new concept for the BB transporter has recently been developed. To properly grip and manipulate each BB segment, the BB transporter has been equipped with a gripper interlock. Due to the geometry of the BB and the vacuum vessel, the attachment point on the BB segment is not aligned with its center of gravity. Hence in addition to the vertical lifting load, large moments about the horizontal axes need to be reacted.The work discussed here concerns the structural analysis conducted on the gripper interlock; its structural integrity has been checked against the most severe load conditions that include also seismic loads according to the EN13001. Elastic analyses were performed using a finite element model in accordance with EN 13001-3-1:2012 + A2:2018, Cranes - General Design - Part 3–1: Limit States and proof competence of steel structure. The effect of the gap sizes at the contact surfaces between gripper interlock and BB after engagement as well as the effect of different friction coefficients on the sliding areas were assessed. The improvements of the design based on the structural analysis are presented, too.

European DEMO Fusion Reactor: Design and Integration of the Breeding Blanket Feeding Pipes

This article describes the design and configuration of the DEMO Breeding Blanket (BB) feeding pipes inside the upper port. As large BB segments require periodic replacement via the upper vertical ports, the space inside the upper port needs to be maximized. At the same time, the size of the upper port is constrained by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. The BB feeding pipes inside the vertical port need to be removed prior to BB maintenance, as they obstruct the removal kinematics. Since they are connected to the BB segments on the top and far from their vertical support on the bottom, the pipes need to be sufficiently flexible to allow for the thermal expansion of the BB segments and the pipes themselves. The optimization and verification of these BB pipes inside the upper port design are critical aspects in the development of DEMO. This article presents the chosen pipe configuration for both BB concepts considered for DEMO (helium- and water-cooled) and their structural verification for some of the most relevant thermal conditions. A 3D model of the pipes forest, both for the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) concepts, has been developed and integrated inside the DEMO Upper Port (UP), Upper Port Ring Channel, and Upper Port Annex (UPA). A preliminary structural analysis of the pipeline was carried out to check the structural integrity of the pipes, their flexibility against the thermal load, their internal pressure, and the deflection induced by the thermal expansion of the BB segments. The results showed that the secondary stress on the hot leg of the HCPB pipeline was above the limit, suggesting future improvements in its shape to increase the flexibility. Moreover, the WCLL concept did not have a critical point in terms of the secondary stress on the pipeline, since the thicknesses and the diameters of these pipes were smaller than the HCPB ones.

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.

Design of the gripper interlock that engages with the DEMO breeding blanket during remote maintenance

The DEMO breeding blanket (BB) must be replaced during the machine lifetime due to the material degradation caused by the neutron irradiation. The large BB segments can therefore be removed through the upper ports of the vacuum vessel by a remotely operated transporter. The size of these ports is however restricted by the magnetic coils causing some of the BB segments to be accessible only on their extremities. The lifting point of these BB segments therefore is away from their centers of gravity also requiring the transfer of bending moments. A concept of the BB transporter was developed recently [1]. It has the required payload capacity and is capable of carrying out also the tilting maneuvers required to extract the BB segments from the VV. The gripper interlock is the interface to the BB segments and is described in this article including the function of its locking mechanism. It has the tightest space constraints of all BB transporter components, and its design is particularly challenging given the large loads to be transferred. The basic concept of the gripper interlock resembles a massive pin with a diameter of approximately 500 mm that is inserted into a countersunk hole in the backside of the BB segment and then locked by an actuated mechanism. The concept allows on the one hand the transfer of large bending moments. The engagement is on the other hand more challenging as compared to the hook of a conventional crane that is required to transfer vertical loads only. In addition the gripper interlock must be designed according to the rules defined for lifting equipment in nuclear power plants and considering increased requirements regarding qualification and in-service inspection since its failure can cause a load drop with the potential to damage the primary confinement.

Development of Mechanical Pipe-Connection Design for DEMO

Maintenance of the DEMO breeding blanket includes the removal and replacement of plasma-facing components. To access the breeding blanket, multiple coolant pipes need to be removed to allow access to the tokamak. As an option to reduce downtime and increase maintenance speed, the pipe-connection concept is developed to allow the removal of multiple pipes at the same time using a remotely operated mechanical connection. The remotely operated multi-pipe Mechanical Pipe Connection (MPC) needs to fulfil multiple requirements, such as high operating temperature and high external forces while at the same time maintaining an acceptable level of sealing between the high-pressure fluid and vacuum surroundings. In addition to the external conditions, the pipes of multiple sizes and fluids are connected in a manifold configuration. Although this will reduce the overall time required to operate the mechanical pipe connection when compared to multiple single-pipe connections, this will introduce additional forces and stresses due the interaction between pipe flow (e.g., simultaneous high- and low-temperature fluid pipes on the same manifold) through the manifold flange. The requirements and the boundary conditions of the multi-pipe MPC are taken into consideration during the design process of MPC. The design process is carried out to find the optimum form and size to allow the mechanical function of the pipe connection during the maintenance phase while withstanding the extreme operating conditions that the MPC will face the during operational phase. The resulting design will then be analyzed using numerical methods to assess the capability of the MPC designs.

Probabilistic failure assessment of an irradiated DEMO breeding blanket component under ferromagnetic loads: Impacts of material data scattering and uncertainty in local stress intensities

The most significant technical challenge in developing structural materials for in-vessel components in the Demonstration Power Plant (DEMO) fusion reactor is to qualify the lifetime of material to guarantee the lifetime of components. The qualification needs to be conducted theoretically rather than empirically because it is conducted based on the knowledge and data acquired in fission neutron irradiation experiments and various simulation irradiation experiments rather than in experiments performed in the actual fusion environment. Such experiments will not be achievable until the DEMO fusion reactor becomes operational. Consequently, various uncertainties are expected in irradiation data and loading conditions.The probabilistic approach, where the probability of failure is calculated based on the probability density function of postulated load distribution and material property distribution, is one path to mitigate such uncertainties. In this approach, those probabilities constitute the subjective probability based on the theoretical understanding of fusion neutron irradiation effects on materials to maximize the confidence in the estimated value.A case study of the Japanese DEMO breeding blanket is the basis for the discussion of the technical issues involved in the probabilistic approach. The breeding blanket design was simplified, the failure probability was estimated, and the impact of electromagnetic force on ferromagnetic material caused by a strong magnetic field induced by toroidal field coils was considered. The discussion of the impact of the irradiation damage is based on the irradiation data for F82H reduced-activation ferritic/martensitic steel. A direction for the development of fusion structural materials is suggested in terms of design technology and the statistical quality of the irradiation database.

Development and application of an alternative modelling approach for the thermo-mechanical analysis of a DEMO water-cooled lithium lead breeding blanket segment

In the frame of the EUROfusion research activities devoted to the design of the DEMO breeding blanket (BB), the Water-Cooled Lithium-Lead BB (WCLL) concept is one of the candidates currently assessed in EU. To this end, an intense research campaign is ongoing to develop a robust geometric configuration for the WCLL BB Central Outboard Segment (COB). Since the current reference design of the WCLL COB segment is not mature enough to allow a full thermal-hydraulic assessment, an alternative procedure aimed at obtaining a thermal field for the whole segment without performing its complete thermal-hydraulic analysis is presented and applied in this work. The scope of the work is to obtain a thermal field for the whole WCLL COB segment detailed enough to ensure a reliable prediction of the secondary stress spatial distribution, which is pivotal for the segment structural analysis.The procedure is articulated in three steps. Firstly, a thermal analysis of the equatorial region of the WCLL COB is performed, in order to calculate the thermal field in the most loaded region of the segment. Then, a multi-region interpolation is carried out in order to obtain a set of polynomial functions of the radial and toroidal variable, accurately representing the thermal field of the assessed equatorial BB slice. Finally, the calculated functions are applied to the whole COB to impose a realistic thermal field to the segment for the prediction of the stress, strain and displacement fields.Since the mesh for the analysis of the whole segment is expected to be coarser than that used for the assessment of its equatorial region, in this paper a mesh independence assessment of the interpolating functions is reported. In particular, their predictive power both in terms of temperature and secondary stress (i. e. thermal-induced stress) is assessed taking into account a decreasing level of detail of the adopted spatial discretization grid.The obtained results, widely reported and discussed hereafter together with the models and the adopted assumptions, show a great predictive power of the adopted polynomial functions, with a good mesh independence of the calculated temperature and stress values.

On the modelling of tritium transport phenomena at fluid-structure interfaces

One of the main functions of the DEMO Breeding Blanket (BB) system is to ensure the tritium breeding inside the reactor. Tritium is a beta emitter radioactive isotope, subjected to several processes that determine its permeation across materials and its leakage towards the environment, posing potential safety issues in terms of radiological hazard. Thus, the evaluation of tritium inventories inside components and tritium losses towards the environment plays a key role in the fulfilment of the pertinent BB safety requirements. In this regard, a research activity has been carried out, in close cooperation between the University of Palermo and the Karlsruhe Institute of Technology, focussing on the development of a multiphysical model that might realistically simulate 3D tritium transport phenomena across complex fluid-structure interfaces. Models, source terms and boundary conditions assumed for the analyses are herewith reported and critically discussed, together with the main results obtained.

Conceptual study of the remote maintenance of the DEMO breeding blanket

The development of a remote maintenance concept to replace DEMO in-vessel components after completion of their lifecycle or in case of failure is fundamental to the successful implementation of the EU fusion roadmap. The replacement of the hot breeding blanket (BB), by far the largest in-vessel component, at the end of its lifecycle is particularly important. This includes the removal from the reactor, the transport to the active maintenance facility (AMF) where the BB is decontaminated and prepared for storage as radioactive waste and the preparation and installation of the new BB. Significant effort is made to control and minimize the spread of contamination. All operations are therefore carried out in sealed rooms and corridors.The high mass of the BB segments requires all remote handling equipment to be capable of handling high payloads of more than 100 tons. It must also operate within tight space and based on impaired feedback from control sensors in the radioactive environment. At the same time, it must be highly reliable in accordance with nuclear requirements and be recoverable in case of failure.Some concepts of BB lifting devices were investigated in the past [1] (Keep et al., 2017), but were discontinued due to insufficient payload capacity. Thus, the vertical maintenance of the BB was identified as one of DEMO's key design integration issues since failure to develop a feasible concept would potentially require major changes to the tokamak architecture [2] (Bachmann et al., 2020). A new study had been initiated with a focus on structural integrity and efficient load transfer from the BB through the RH equipment to the VV upper port. A concept of the BB transfer cask and the BB transporter resulting from this study is presented in this article together with a conceptual study of the layout of the tokamak building and the AMF. Studies of alternative concepts for in-vessel maintenance are conducted in parallel but will not be described here.