The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have allowed a quite advanced reactor architecture to be achieved. Moreover, significant efforts have been made in order to develop the WCLL BB pre-conceptual design following a holistic approach, identifying interfaces between components and systems while respecting a system engineering approach. This paper reports a description of the current WCLL BB architecture, focusing on the latest modifications in the BB reference layout aimed at evolving the design from its pre-conceptual version into a robust conceptual layout. In particular, the main rationale behind design choices and the BB’s overall performances are highlighted. The present paper also gives an overview of the integration between the BB and the different in-vessel systems interacting with it. In particular, interfaces with the tritium extraction and removal (TER) system and the primary heat transfer system (PHTS) are described. Attention is also paid to auxiliary systems devoted to heat the plasma, such as electron cyclotron heating (ECH). Indeed, the integration of this system in the BB will strongly impact the segment design since it envisages the introduction of significant cut-outs in the BB layout. A preliminary CAD model of the central outboard blanket (COB) segment housing the ECH cut-out has been set up and is reported in this paper. The chosen modeling strategy, adopted loads and boundary conditions, as well as obtained results, are reported in the paper and critically discussed.

The EU-DEMO must demonstrate the possibility of generating electricity through nuclear fusion reactions. Moreover, it must denote the necessary technologies to control a powerful plasma with adequate availability and to meet the safety requirements for plant licensing. However, the extensive radioactive materials inventory, the complexity of the plant, and the presence of massive energy sources require a rigorous safety approach to fully realize fusion power’s environmental advantages. The Tokamak building barrier design must address two main issues: radioactive mass transport hazards and energy-related or pressure/vacuum hazards. Safety studies are performed in the frame of the EUROfusion Safety And Environment (SAE) work package to support design improvement and evaluate the thermal-hydraulic behavior of confinement building environments during accident conditions in addition to source term mobilization. This paper focuses on developing a thermal-hydraulic model of the EU-DEMO Tokamak building. A preliminary model of the heat ventilation and air conditioning system and vent detritiation system is developed. A loss-of-coolant accident is studied by investigating the Tokamak building pressurization, source term mobilization, and release. Different nodalizations were compared, highlighting their effects on source term estimation. Results suggest that the building design should be improved to maintain the pressure below safety limits; some mitigative systems are preliminarily investigated for this purpose.

In the broad framework of the nuclear power plants (NPPs) industry, the dynamic probabilistic risk assessment could answer the time dependence deficiency of the event tree and fault tree analysis. The basic event tree approach relies on experts' pre-constructed accident sequences without exploring the time-dependent nature of an accident scenario, which could strongly affect the accident sequence. Conversely, the effects of events timing can be studied by adopting a dynamic event tree (DET) approach. Developing a DET methodology requires integrating a system code capable of replicating an accident scenario and a logic-driver code able to generate the event tree sequence, trigger plant safety systems, and manage other relevant events throughout the simulation. For this purpose, Methods for Estimation of Leakages and Consequences of Release (MELCOR) and reactor analysis and virtual control environment (RAVEN) have been coupled through a Python script developed by the Sapienza University of Rome, Rome, Italy, to perform DET studies during accident transients in fusion and fission reactors. RAVEN is a software tool developed at the Idaho National Laboratory (INL), Idaho Falls, ID, USA, to act as a control logic driver and post-processing tool for different applications. MELCOR for fusion is a fully integrated design basis and severe accident code that simulates thermal-hydraulic behavior and self-consistently accounting for aerosol transport in nuclear facilities and reactor cooling systems for the evaluation of the source term in fusion reactors. The coupling between these codes will provide a wide range of NPP risk assessment analyses, establishing new best practices. In this work, a preliminary DET study has been performed, selecting as initiating event an ex-vessel loss of coolant accident (LOCA) in the water-cooled lithium lead test blanket system (WCLL TBS) to be tested in International Thermonuclear Experimental Reactor (ITER). Time-dependent parameters such as the intervention of the plasma shutdown system and the closure of the main system isolation valves have been sampled to study evolving system scenarios.

As part of the pre-conceptual design activities for the European DEMOnstration plant, a carefully selected set of safety analyses have been performed to assess plant integrated performance and the capability to achieve expected targets while keeping it in a safe operation domain. The DEMO divertor is the in-vessel component in charge of exhausting the major part of the plasma ions’ thermal power in a region far from the plasma core to control plasma pollution. The divertor system accomplishes this goal by means of assemblies of cassette and target plasma facing units modules, respectively cooled with two independentheat-transfer systems. A deterministic assessment of a divertor in-vessel Loss-of-Coolant Accident is here considered. Both Design Basis Accident case simulating the rupture of an in-vessel pipe for the divertor cassette cooling loop, and a Design Extension Conditions accident case considering the additional rupture of an independent divertor target cooling loop are assessed. The plant response to such accidents is investigated, a comparison of the transient evolution in the two cases is provided, and design robustness with respect to safety objectives is discussed.