Embrittlement of WCLL blanket and its fracture mechanical assessment

In the European fusion programme, the Water Cooled Lithium Lead breeding blanket (WCLL BB) uses EUROFER as a structural material cooled with water at temperatures between 295 °C–328 °C and a pressure of 155 bar. The WCLL BB will be significantly irradiated (>2 dpa), while some parts will not receive significant heat loads, e.g. the sidewalls or the back-supporting structures. The irradiation, together with the irradiation temperature of EUROFER below 350 °C, produces a shift of the ductile-to-brittle-transition temperature (DBTT) to levels above room temperature at neutron doses, causing material damage as low as 2–3 dpa. Even though the DBTT does not reach the operating temperature level, brittle/non-ductile fracture is a concern during in-vessel maintenance when the BB temperature is below the DBTT. Two loading scenarios were identified as severe in this respect: (i) re-pressurization of the WCLL BB cooling loop after in-vessel maintenance, and (ii) dead weight loads during lifting of the BB segment. The embrittlement of the WCLL BB was investigated by quantifying the local DBTT shift in its parts based on current knowledge of the embrittlement behaviour of EUROFER under neutron irradiation. Therefore, a suitable, not overly conservative procedure was derived considering dpa damage and transmuted helium effects. The results demonstrate the ability to identify the 3D spread of the severely embrittled zones in the structure whose impact on the structural integrity was assessed considering the risk of brittle/non-ductile fracture. Thereby, the fracture mechanics approach established in nuclear codes was applied assuming its applicability to EUROFER. The embrittled zones in the first wall (FW) and its sidewalls pass the criteria when assessing the relatively low stresses resulting from the coolant pressure. The assessment was then continued considering stresses appearing in the FW during maintenance, in particular, when lifting the BB segment and transporting it out of the vacuum vessel. In this context, the maximum tolerable flaw sizes were determined in a parameter study considering designs of the FW with different cooling channel wall thicknesses

DEMO WCLL BB breeding zone cooling system design: analysis and discussion

The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium self-sufficiency, shielding the Vacuum Vessel and removing the heat generated in the tokamak plasma. The last function is fulfilled by the First Wall (FW) and Breeding Zone (BZ) independent cooling systems. Several layouts of BZ coolant system have been investigated in the last years in order to identify a configuration that guarantee Eurofer temperature below the limit (823 K) and good thermal-hydraulic performances (i.e. water outlet temperature 601 K). A research activity is conducted to study and compare four configurations, which rely on different arrangement of the stiffening plates (i.e. toroidal-poloidal and radial-poloidal), orientation of the cooling pipes (i.e. horizontal, vertical) and PbLi flow path. The analysis is carried out using a CFD codes, thus a threedimensional finite volume model of each configuration is developed, adopting the commercial ANSYS CFX code. The objective is to compare the BZ cooling system layouts, identifying and discussing advantages and key issues from the thermal-hydraulic point of view, also considering feedbacks from MHD and neutronics analyses. The research activity aims at laying the groundwork for the finalization of the WCLL blanket design, pointing out relevant thermal-hydraulic aspects

Model development and transient analysis of the wcll bb bop demo configuration using the apros system code

Extensive modelling and analytical work has been carried out considering the water-cooled lithium–lead breeding blanket (WCLL BB) balance of plant (BOP) configuration of the demonstration power plant (DEMO) using the Apros system code, developed by VTT Technical Research Centre of Finland Ltd. and Fortum. Contributing to the BOP work package of the EUROfusion Consortium, the integral plant model for dynamic analyses of the WCLL BB configuration has been updated with special attention to primary system components. Following trends of relevant neutronics modelling, a new BB model has been implemented in 2020 with the aim to obtain higher resolution output data and a more realistic thermalhydraulic feedback from the primary system. Once-through steam generator user components have been built based on CAD models conceived by BOP partners. Transient analyses have been performed providing a better picture regarding the behaviour of main components, e.g., the BB and the OTSGs, whilst highlighting possible ways to optimise the control scheme of the plant

Parametric thermal analysis for the optimization of Double Walled Tubes layout in the Water Cooled Lithium Lead inboard blanket of DEMO fusion reactor

Within the roadmap that will lead to the nuclear fusion exploitation for electric energy generation, the construction of a DEMOnstration (DEMO) reactor is, probably, the most important milestone to be reached since it will demonstrate the technological feasibility and economic competitiveness of an industrial-scale nuclear fusion reactor. In order to reach this goal, several European universities and research centres have joined their efforts in the EUROfusion action, funded by HORIZON 2020 UE programme. Within the framework of EUROfusion research activities, ENEA and University of Palermo are involved in the design of the Water-Cooled Lithium Lead Breeding Blanket (WCLL BB), that is one of the two BB concepts under consideration to be adopted in the DEMO reactor. It is mainly characterized by a liquid lithium-lead eutectic alloy acting as breeder (lithium) and neutron multiplier (lead), as well as by subcooled pressurized water as coolant. Two separate circuits, both characterized by a pressure of 15.5 MPa and inlet/outlet temperatures of 295 °C/328 °C, are deputed to cool down the First Wall (FW) and the Breeder Zone (BZ). The former consists in a system of radial-toroidal-radial C-shaped squared channels where countercurrent water flow occurs while the latter relies in the use of bundles of poloidal-radial Double Walled Tubes (DWTs) housed within the breeder. A parametric thermal study has been carried out in order to assess the best DWTs' layout assuring that the structural material maximum temperature does not overcome the allowable limit of 550 °C and that the overall coolant thermal rise fulfils the design target value of 33 °C. The study has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the quoted Abaqus FEM code. Main assumptions and models together with results obtained are herewith reported and critically discussed

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

Design of the Test Section for the Experimental Validation of Antipermeation and Corrosion Barriers for WCLL BB

Tritium permeation into the Primary Heat Transfer System (PHTS) of DEMO and ITER reactors is one of the challenging issues to be solved in order to demonstrate the feasibility of nuclear fusion power plants construction. Several technologies were investigated as antipermeation and corrosion barriers to reduce the tritium permeation flux from the breeder into the PHTS. Within this frame, alumina coating manufactured by Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) are two of the main candidates for the Water Cooled Lithium Lead (WCLL) Breeder Blanket (BB). In order to validate the performance of the coatings on relevant WCLL BB geometries, a mock-up was designed and will be characterized in an experimental facility operating with flowing lithium-lead, called TRIEX-II. The present work aims to illustrate the preliminary engineering design of a WCLL BB mock-up in order to deeply investigate permeation of hydrogen isotopes through PHTS water pipes. The permeation tests are planned in the temperature range between 330 and 500 °C, with hydrogen and deuterium partial pressure in the range of 1–1000 Pa. The hydrogen isotopes transport analysis carried out for the design and integration of the mock-up in TRIEX-II facility is also shown

Thermomechanical and Thermofluid-Dynamic Coupled Analysis of the Top Cap Region of the Water-Cooled Lithium Lead Breeding Blanket for the EU DEMO Fusion Reactor

In the EU, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept is one of the candidates for the design of the DEMO reactor. From the past campaign of analysis emerged that the thermal-induced stress led to the failure in the verification of the RCC-MRx structural criteria. Hence, in this paper the classic conceptual design approach, based on a pure FEM thermal and structural analysis, is compared to a coupled thermofluid-dynamic/structural one. Even though the coupled approach requires tremendous modelling effort and computational burden, it surely allows determining the thermal field with a higher level of detail than the FEM analysis. Therefore, in this work, the focus is put on the impact of a more detailed thermal field on the DEMO WCLL BB global structural performances, focusing on the Top Cap region of its Central Outboard Blanket segment. The obtained results have allowed confirming the soundness of the design solution of the Top Cap region, except for concerns arising on the mass flow rate distribution. Moreover, results have shown that, globally, the pure FEM approach allows for obtaining more conservative results than the coupled one. This is a positive outcome in sight of the follow-up of the DEMO WCLL BB design, as it will be still possible adopting the pure FEM approach to quickly down-select design alternatives, using the most onerous coupled approach to finalise the most promising

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 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

Multiphysics Optimization for First Wall Design Enhancement in Water-Cooled Breeding Blankets

The commercial feasibility of the first fusion power plant generation adopting D-T plasma is strongly dependent upon the self-sustainability in terms of tritium fuelling. Within such a kind of reactor, the component selected to house the tritium breeding reactions is the breeding blanket, which is further assigned to heat power removal and radiation shielding functions. As a consequence of both its role and position, the breeding blanket is heavily exposed to both surface and volumetric heat loads and, hence, its design requires a typical multiphysics approach, from the neutronics to the thermo-mechanics. During last years, a great deal of effort has been put in the optimization of the breeding blanket design, with the aim of maximizing the tritium breeding and heat removal performances without undermining its structural integrity. In this paper, a derivative-free optimization method named “Complex method” is applied for the design optimization of the European DEMO Water-Cooled Lithium Lead breeding blanket concept. To this purpose, a potential performances-based objective function, focusing on the maximization of the tritium breeding, is defined and a multiphysics numerical model of the blanket is developed in order to solve the coupled thermo-mechanical problem, while the optimization algorithm leads the design towards a minimum optimum point compliant with the prescribed requirements. Once the optimized design is obtained, its nuclear and thermo-structural performances are assessed by means of specific neutron transport and multiphysics simulations, respectively. Finally, the structural integrity is verified by means of the application of the RCC-MRx design criteria