On the thermal dynamic behaviour of the helium-cooled DEMO fusion reactor

The EU-DEMO conceptual design is being conducted among research institutions and universities from 26 countries of European Union, Switzerland and Ukraine. Its mission is to realise electricity from nuclear fusion reaction by 2050. As DEMO has been conceived to deliver net electricity to the grid, the choice of the Breeding Blanket (BB) coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. In particular, due to the pulsed nature of the heat source, the Primary Heat Transfer System (PHTS) becomes a very important actor of the Balance of Plant (BoP) together with the Power Conversion System (PCS). Moreover, aiming to mitigate the potential negative impact of plasma pulsing on BoP equipment, for the DEMO plant is also being investigated a "heat transfer chain" option which envisages an Intermediate Heat Transfer System (IHTS) equipped with an Energy Storage System (ESS) between PHTS and PCS. Within this framework, a preliminary study has been carried out to analyse the thermal dynamic behaviour of the IHTS system for the Helium-Cooled Pebble Bed (HCPB) BB concept during pulse/dwell transition which should be still considered as the normal operating mode of a fusion power plant. Starting from preliminary thermal-hydraulic calculations made in order to size the main BoP components, the global performances of DEMO BoP have been quantitatively assessed focusing the attention on the attitude of the whole IHTS to smooth the sudden power variations which come from the plasma. The paper describes criteria and rationale followed to develop a numerical model which manages to simulate simple transient scenarios of DEMO BoP. Results of numerical simulations are presented and critically discussed in order to point out the main issues that DEMO BoP has to overcome to achieve a viable electricity power output

Investigation of fatigue damage growth and self-heating behaviour of cross-ply laminates using simulation-driven dynamic test

Structural integrity of aerospace assets is paramount for both the safety and economy of aviation industry. The introduction of composites into the design of aero-structures generated several economic benefits but also led to several challenges, including fatigue damage growth and self-heating behaviour. Fatigue of metals is widely managed by calculations of damage accumulation and prediction of residual life. These techniques do not always apply to the fatigue of composites, where the onset and propagation of damage are still under investigation. Furthermore, vibration-induced fatigue is even less understood because of a handful of failure criteria available and, also, because it is biased by the self-heating conditions of the material itself. The authors have underpinned one failure criterion for vibration fatigue and mapped that against self-heating and environmental temperatures. Despite the advances, several research questions were left open because of the complex multiphysics behaviour of fatigue which outreached the experimental capacity. Therefore, this research suggests a Simulation-Driven Dynamic Test (SDDT) framework that deconstructs vibration fatigue experiments into step-wise steady-state analyses. This novel approach will enable (a) investigating the failure mode mixity of the underlying failure criterion, and (b) simulating the surface temperature during the delamination growth under vibration conditions.</p

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

Conceptual design of the enhanced coolant purification systems for the European HCLL and HCPB test blanket modules

The Coolant Purification Systems (CPSs) is one of the most relevant ancillary systems of European Helium Cooled Lead Lithium (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Modules (TBMs) which are currently in the preliminary design phase in view of their installation and operation in ITER. The CPS implements mainly two functions: the extraction and concentration of the tritium permeated from the TBM modules into the primary cooling circuit and the chemistry control of helium primary coolant. During the HCLL and HCPB-TBSs (Test Blanket Systems) Conceptual Design Review (CDR) in 2015 it was recognized the need of reducing the tritium permeation into the Port Cell #16 of ITER. To achieve this and, then, to lower the tritium partial pressure in the Helium Cooling Systems in normal operation, the helium flow-rate treated by each CPS has been increased of almost one order of magnitude. In 2017, to satisfy the CDR outcomes and the new design requirements requested by Fusion for Energy (F4E, the European Domestic Agency for ITER), ENEA performed a preliminary design of the “enhanced” CPSs. This paper presents the current design of the “enhanced” CPSs, focusing on design requirements, assumptions, selection of technologies and preliminary components sizing

Structural assessment of the EU-DEMO water-cooled lead lithium central outboard blanket segment adopting the sub-modelling technique

The development of a sound conceptual design of the Water-Cooled Lead Lithium Breeding Blanket (WCLL BB) is pivotal to make a breakthrough towards the selection of the driver blanket concept for the EU-DEMO. To achieve this goal, an intense research campaign has been performed at the University of Palermo, in cooperation with ENEA Brasimone, under the umbrella of EUROfusion. In this paper, structural analyses of different poloidal regions of the WCLL BB Central Outboard Blanket (COB) segment are reported. In particular, starting from the results of the thermo-mechanical analysis of the whole WCLL BB COB segment, the sub-modelling technique has been applied to the most significant poloidal regions, located at the top, middle and bottom of the segment. The aim is to focus on the stress field locally arising under purposely selected steady-state nominal and accidental loading scenarios. The nominal BB operating conditions, as well as steady-state scenarios derived from both the in-box LOCA and Vertical Plasma Disruption accidents have been considered. Thanks to the sub-modelling approach, the deformative action of the entire segment can be imposed at the boundaries of each local model to realistically assess its structural performances. Moreover, each local model reproduces structural details not included in the global one, such as the Segment Box (SB) cooling channels. Then, the structural behaviour of the selected regions has been assessed in compliance with the RCC-MRx code. The obtained results highlighted that the structural behaviour predicted by the whole segment analysis is similar to that predicted by sub-modelling calculations within the Stiffening Plates, whereas the application of the sub-modelling is a must to investigate in detail the SB structural performances. In addition, results indicate that the BB attachments should be revised, as they contribute to produce the WCLL COB large deformation originating excessive stresses, mainly within the equatorial region

Development of a thermal-hydraulic model of the EU-DEMO Water Cooled Lithium Lead Breeding Blanket Primary Heat Transport System

The EUROfusion consortium is developing the project of a DEMOnstration Fusion Reactor (EU-DEMO) which would follow ITER in the pathway towards the quest for the exploitation of fusion energy. EU-DEMO has been conceived to deliver net electric power to the grid. Therefore, proper critical evaluations of the tokamak cooling and power conversion systems are needed because they play a pivotal role in the design and licencing of the overall plant. The EU-DEMO reactor will be based on the tokamak concept and, as such, it is supposed to undergo a pulsed duty cycle under normal conditions, which might challenge the qualified lifetime of the main equipment inducing undue thermal and mechanical cycling. Moreover, the EU-DEMO plasma control strategy postulates the possible occurrence of planned and off-normal plasma overpower transients that might jeopardise the structural integrity of the plasma facing components. It is, therefore, of paramount importance to have appropriate tools to reproduce the thermal-hydraulic behaviour of tokamak cooling systems during major operational and accidental scenarios in a realistic and reliable way. In this context, University of Palermo in cooperation with EUROfusion has developed a finite volume model of the Primary Heat Transport System (PHTS) feeding the EU-DEMO Water Cooled Lithium Lead Breeding Blanket (WCLL BB). The activity has been led following a theoretical–computational approach based on the adoption of the TRACE thermal-hydraulic system code. Particular attention has been paid to capturing all the main geometrical, hydraulic and heat transfer features characterising both in-vessel and ex-vessel components. Preliminary analyses have also been carried out to check the code's predictive potential in fusion relevant applications. Models, assumptions, and outcomes of the analyses are herewith reported and critically discussed

Exploratory Thermo-Mechanical Assessment of the Bottom Cap Region of the EU DEMO Water-Cooled Lead Lithium Central Outboard Blanket Segment

The Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB) is one of the two BB concept candidates to be selected as the driver blanket for the EU DEMO fusion reactor. In this regard, the development of a sound architecture of the WCLL Central Outboard Blanket (COB) Segment, ensuring the fulfilment of the thermal and structural design requirements, is one of the main goals of the EUROfusion consortium. To this purpose, an exploratory research campaign has been launched to preliminarily investigate the thermo-mechanical performances of the Bottom Cap (BC) region of the WCLL COB segment because of its peculiarities making its design different from the other regions. The assessment has been carried out considering the nominal BB operating conditions, the Normal Operation (NO) scenario, as well as a steady-state scenario derived from the in-box LOCA accident, the Over-Pressurization (OP) scenario. Starting from the reference geometric layout of the WCLL COB BC region, a first set of analyses has been launched in order to evaluate its structural performances under a previously calculated thermal field and to select potential geometric improvements. Then, the analysis of a complete BC region was conducted from both the thermal and structural standpoints, evaluating its structural behaviour in compliance with the RCC-MRx code. Finally, after some iterations and geometric updates, a promising geometric layout of the BC region has been obtained even though some criticalities still persist in the internal Stiffening Plates and First Wall. However, the obtained results clearly showed that the proposed layout is worthy to be further assessed to achieve a robust enough configuration. The work has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the quoted Ansys commercial FEM code