EU DEMO Research Articles

Automated maintenance feasibility testing on the EU DEMO Automated Inspection and Maintenance Test Unit (AIM-TU)

In order to reach commercially-relevant availability, future fusion power plants must minimise the duration of maintenance shutdowns. However, as radiation levels increase, so too will the number of maintenance tasks that must be performed without human access. To meet these conflicting constraints, remote maintenance systems must therefore significantly increase their capabilities, performing tasks faster and in parallel. Unfortunately, the current teleoperation-based maintenance approach is inherently limited in these regards. Automation emerges as a potential solution, having demonstrated dramatic productivity gains in manufacturing across the world. The European DEMO project has begun to explore the feasibility of automated maintenance as a route to meeting the challenging availability targets. To this end, the Automated Inspection and Maintenance Test Unit (AIM-TU) has been designed and constructed, providing a highly versatile test platform able to adapt to evolving R&D programmes. It consists of a modular robot cell equipped with two robot arms mounted on linear rails, interchangeable end-of-arm-tooling, a variety of vision sensors and a fully integrated safety system. The hardware is supported by a modular software architecture, which permits different control input modes and provides a digital twin simulation of the cell for virtual verification and validation of control algorithms. This paper provides an overview of the system’s capabilities and reports on the first automated maintenance feasibility tests performed on the cell: automatic replacement of JET reactor tiles and automatic contactless inspection of tile anomalies.

Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger

In the framework of the EU DEMO fusion reactor technology development, a new steam generator consisting in a helical tube bundle is currently under study for the lithium-lead loop of the Dual Coolant Lithium Lead (DCLL) and Water Cooled Lithium Lead (WCLL) breeding blankets. This solution turns out to be very interesting for high power plants, since the helical geometry is very compact and it assures a high thermal power exchanged, taking up the minimum amount of space.In this framework, the ENEA Brasimone Research Centre supports the development of this innovative component by means of experimental activities, involving CIRCE, a large scale pool-type facility using Lead-Bismuth Eutectic as primary fluid and pressurized water as secondary fluid. The main components of a new test section named THETIS are: a Fuel Pin Simulator, acting as primary heating source, a vertical mechanical pump for the primary fluid circulation inside the main pool and a new prototypical helical coil steam generator mock-up, designed to be relevant for the DCLL and WCLL lithium-lead loop, acting as primary heat sink. In such configuration, the facility will be involved in a set of tests aiming at demonstrating the technological feasibility and thermal-hydraulic performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. The experiments will also provide a database for system thermal-hydraulic codes validation.The aim of this paper is to present the main layout of the CIRCE facility, to describe the preliminary design of the test section and the main features of the helical coil steam generator mock-up. Furthermore, a preliminary test analysis carried out by the system thermal hydraulic code RELAP5/Mod3.3 is presented. A numerical 1-D model of the helical coil steam generator has been set-up in order to test the performance of the component from a thermal-hydraulic point of view. An additional 3D CFD analysis was performed for the flow field of the HCSG to assess the secondary flow behaviour in the component and the pressure losses.

Pre-conceptual design of EU DEMO balance of plant systems: Objectives and challenges

The European Research Roadmap to the Realisation of Fusion Energy foresees that the DEMO reactor is going to succeed ITER in the pathway towards the exploitation of nuclear fusion, achieving long plasma operation time, demonstrating tritium self-sufficiency and producing net electric output on an industrial scale. Therefore, its design must be more oriented towards the Balance of Plant (BoP) than it is in ITER.Since the early pre-conceptual phase of the DEMO project, emphasis has been laid on identifying the main requirements affecting the overall architecture of the BoP. For instance, specific efforts and proper solutions have been envisaged to cope with the pulsed nature of the heat source. Furthermore, the current development of two blanket concepts calls for two separate BoP options to be conceived.This paper summarizes the main alternatives outlined at the end of DEMO pre-conceptual design phase for the BoP concepts based on both the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) technologies. Then, the assumed reference configurations of both the BoP concepts are described in detail, highlighting the main features and the most relevant engineering aspects. Attention will be focussed on technological challenges, integration constraints and other open issues, highlighting pros and cons of the chosen BoP options to be further investigated in the next design phase.

Development of a PbLi heat exchanger for EU DEMO fusion reactor: Experimental test and system code assessment

In the framework of the DEMO Work Package Balance of Plant of EUROfusion consortium, ENEA has been involved in experimental and numerical activities related to the development of a prototypical heat exchanger, suitable as steam generator for the lithium-lead loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. For this purpose, an experimental campaign has been executed on the pool-type liquid metal-cooled facility CIRCE at ENEA Brasimone Research Centre.A dedicated test section named HERO has been designed and installed inside the main vessel of CIRCE. The innovative steam generator consists of a tube bundle with seven double walled bayonet tubes, fed with pressurized water. The selected configuration improves the plant safety, reducing the possibility of water-lead/lead-alloy interaction thanks to a double physical separation and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas.A set of tests has been defined to demonstrate technological feasibility and performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. In particular, one of the performed tests is presented and discussed in this paper. The experiment is characterized by a secondary loop feedwater working pressure of 10 MPa and a steam generator inlet temperature of 300 °C. On the primary side, the lead-bismuth eutectic has been used as working fluid with a steam generator inlet temperature of 480 °C. During the test, an experimental sensitivity analysis on the primary coolant mass flow rate has been performed. Furthermore, the results of a post-test analysis realized with two versions of the system thermal-hydraulic code RELAP5 are presented, in order to evaluate their capability in simulating the performances of the component and to support the validation process of the codes for heavy liquid metal applications. The work is concluded presenting a scaling analysis to find the equivalence between LBE and PbLi, recalculating the available experimental data with RELAP5 code using PbLi as working fluid.

Comparison between finite element and experimental evidences of innovative W lattice materials for sacrificial limiter applications

Power exhaust is a key mission for the realization of fusion electricity. Engineering challenges may arise from the extreme heat fluxes developed during plasma transients, above the limit offered by existing materials. These can reduce the lifetime of plasma-facing components (PFCs), imposing extraordinary maintenance, reactor safety issues and ultimately delayed return to normal operation. Concerning the EU DEMO reactor, discrete sacrificial limiters are being investigated as the last safety resource of the reactor’s wall in case of unmitigated events. Within this context, micro-engineered tungsten (W) lattices are proposed to cope with unmitigated plasma disruptions. Unlike bulk W, lattices can be tailored to meet the operational requirements of the limiter, compromise between steady-state and off-design performances while avoiding overloading of the heat sink and delay the need for extraordinary maintenance. By calibrating an equivalent solid model originally developed and validated for open-cell aluminum (Al) foams, tailored lattices have been modelled and samples fabricated through additive manufacturing for characterization and testing, currently ongoing.In the present work, the thermal response of lattice samples during thermal shock high heat flux (HHF) tests performed at the linear facility QSPA Kh-50 facility is simulated using ANSYS and compared with available results. Enthalpy changes of W were imposed to simulate phase change. Good agreement with experiments and SDC-IC reference up to melting point was observed. Ultimately, a thermal quench of an unmitigated DEMO disruption was simulated involving an original MAPDL routine that removes mesh elements at the melting or vaporization point.

Conceptual design of a PAV-based tritium extractor for the WCLL breeding blanket of the EU DEMO: Effects of surface-limited vs. diffusion-limited modeling

In order to close the fuel cycle, the EU DEMO will be equipped with a breeding blanket (BB) for the on-site tritium production. In the water cooled lithium lead (WCLL) BB concept, the tritium will be generated in the liquid PbLi from which it must then be extracted.The conceptual design of the tritium extraction and removal system for this BB concept is addressed here. It relies on the permeator against vacuum (PAV) technology, where the tritium dissolved in the PbLi is forced to cross a metal (Nb) membrane by pumping the vacuum on the secondary side. The proposed PAV geometrical configuration is based on the shell-and-tube concept, where the walls of the pipes in which the PbLi flows are the permeator membrane. The dimensioning of one permeator vessel for one of the PbLi loops is carried out here, adopting two different models for the tritium permeation: diffusion-limited (for the case of a non-contaminated membrane) and surface-limited (suitable for a contaminated Nb surface).The results show that in the case of a non-contaminated membrane it is possible to reach the target tritium extraction efficiency also at the PbLi nominal operating temperature, while in the case of a contaminated membrane the PbLi should be heated to enhance the permeation.

Integration of LiPb loops for WCLL BB of European DEMO

The eutectic alloy Lithium Lead (LiPb) enriched at 90 % in 6Li is the breeder material for one of the candidate European Breeding Blanket (BB) concepts. Currently under investigation for DEMO reactor, the Water Cooled Lithium-Lead (WCLL), and for the WCLL Test Blanket Module (TBM) that will be qualified in the ITER reactor. The LiPb alloy is used as tritium breeder, neutron multiplier and tritium carrier. The design of the LiPb loops is currently under study and the conceptual design of the main loop components has been completed. For this reason, it becomes mandatory to proceed with the integration of the LiPb loops in the EU DEMO Tokamak building, checking the consistency of the different systems design to be integrated in DEMO reactor building. CAD design and integration of the entire LiPb loops are shown taking into account the building areas assigned, the interfaces with the other systems and the requirement related to the LiPb loop functions. An initial layout of the pipework and the position of the main components have been defined on the basis of the following design requirements: (I) gamma radiation shielding of the components and the pipework; (II) target flow velocity of the LiPb; (III) thermal expansion of the pipes; (IV) possibility to drain the entire loop; (V) redundancy of the loops; (VI) remote maintenance; (VII) position in the building and dimensions of the storage tanks. The 3D model of the entire loops has been provided and integrated in DEMO Tokamak building pointing out the issues related to the interfaces with the other systems and with the building itself.

Alternative PF Coil Winding Pack Design for the EU DEMO

The EUROfusion DEMO machine is at its pre-conceptual stage. The design of the magnet system for this machine is being developed in collaboration with several European research centers. The Poloidal Field (PF) coils proposed by the Swiss Plasma Center (SPC) use a multiple-in-hand Pancake winding technique. A design based on updated plasma scenarios (2018) was recently proposed. The Winding Pack (WP) aspect ratio was kept close to one, similar to the baseline design in the EUROfusion 2018 DEMO baseline document. In this publication, we study the effects of changing WP aspect ratio. A vertically elongated WP cross-section is beneficial in the reduction of the peak magnetic field at WP cross-section, the machine radial extent, and the hydraulic lengths. However, it increases the bending stress on the WP. An updated PF design is presented. The dimensioning and allocation of materials is based on 3-dimensional electromagnetic calculations. Mechanical performance is evaluated against both dynamic fatigue stress and static stresses. The static stresses are evaluated through 3-dimensional finite element simulations to take into account also the ripple effect of TF coils and increase in bending stresses due to the elongation of the WP cross-section. The thermal-hydraulic analysis verifies a temperature margin of 1.5 K and confirms sufficient coolant flow for the prescribed pressure drop. The quench analysis determines the hot-spot temperatures during various quench scenarios.

CFD analysis of natural convection cooling of the in-vessel components during a shutdown of the EU DEMO fusion reactor

In view of the large neutron fluence expected in a fusion power plant, the maintenance of the in-vessel components (IVC) must be carried out using Remote Handling (RH); however, before the RH robots can intervene, the temperature of the IVCs must be reduced, so a cooldown phase is required after the reactor shutdown before maintenance activities can start. In the EU DEMO two options are being investigated to cool down the Breeding Blanket (BB) structures before maintenance, namely introducing fans to pump air in forced convection in the plasma chamber (after opening the Vacuum Vessel), or letting the air at room temperature cool down the structures by natural convection; if the required downtime is acceptable, the second option is clearly preferred, as it would reduce the cost and complexity of the system. This work analyses the natural convection option via a 3D transient Computational Fluid-Dynamics (CFD) conjugate heat transfer model, to evaluate the required time to cool down the BB.

Experimental Investigation on CIRCE-HERO for the EU DEMO PbLi/Water Heat Exchanger Development

The present paper describes the experimental campaign executed at the ENEA Brasimone Research Centre aiming at supporting the development of a PbLi/water heat exchanger suitable for the lithium–lead loops of the dual coolant lithium lead and the water cooled lithium lead breeding blankets of the EU DEMO fusion reactor. The experiments were performed in a test section named HERO, installed inside the main vessel of the lead–bismuth eutectic-cooled pool-type facility CIRCE. The test section hosts a steam generator bayonet tube mock-up in relevant scale, which was selected as a promising configuration for DEMO purposes. For the thermal-hydraulic characterization of the component, five tests were executed at different water pressures (6, 8, 12 MPa, two tests at 10 MPa), and liquid metal flow rates (40, 33, 27, 20, 10 kg/s). The experimental outcomes proved the technological feasibility of this novel steam generator and its suitability for the DEMO PbLi loops. The activity was completed with a post-test analysis using two versions of the system code RELAP5. Because the experiments were executed with lead–bismuth eutectic, a scaling analysis is proposed to find the equivalence with PbLi. RELAP5 code was applied to recalculate the experimental data using PbLi as working fluid.

Current status and future perspectives of EU ceramic breeder development

• EU reference tritium breeding pebbles are produced using a melt-based process. • Process scale-up is underway to ensure the supply for ITER. • Biphasic materials show enhanced mechanical strength and stability. • Design relevant data were summarised in a material property handbook. • Crucial neutron irradiation experiments will complement the material qualification Biphasic ceramic pebbles, consisting of Li 4 SiO 4 and Li 2 TiO 3 , are being developed as the EU reference tritium breeding material for ITER and DEMO. A modified melt-based process for the pebble fabrication was established and optimised with regard to the process stability and yield. In parallel, the properties of pebbles and pebble beds were evaluated in dedicated experiments. These include investigations into e.g. the long-term thermal stability of pebbles, the thermomechanical properties of pebble beds, as well as the compatibility between the ceramics and the structural material. The available data was recently summarised and issued as the Material Property Handbook on Advanced Ceramic Breeders. During the EU DEMO conceptual phase, two main issues are proposed to be pursued in view of ITER. Irradiation campaigns are planned to ensure that the advanced ceramic breeder material is adequately tested and an extensive process scale-up with an increased production rate is foreseen to secure the supply for the ITER test blanket modules.

Pressure suppression system influence on vacuum vessel thermal-hydraulics and on source term mobilization during a multiple first Wall – Blanket pipe break

Being the Vacuum Vessel Pressure Suppression System (VVPSS) one of the most important passive safety systems to be foreseen in DEMO plant, design and integration challenges have to be faced to ensure that best performance within safety requirements are always achieved. In this framework, parametric safety analyses have been performed to support VVPSS design activities; in particular to determine the minimum flow area required by the suppression system pipework to limit the vacuum vessel pressure below the limit imposed as a requirement by design. The selected Postulated Initiating Event (PIE) is a double-ended guillotine break in the Primary Heat Transfer System (PHTS) feeding pipe of the Breeding Zone (BZ) during plasma activity. Coolant is discharged inside the VV upper port volume and an unmitigated disruption occurs, causing further breaks in the first wall (FW) cooling channels. Considering that limiters could be introduced in the future design of the EU DEMO reactor to prevent damages to plasma-facing components, the same parametric study has been performed considering limiters accident mitigation effects. Because discharge flow area toward the suppression pool could also affect the source terms mobilization, the transport of radioactive products (e.g. tritium, tungsten dust and activated corrosion products) has been simulated with the fusion version of the MELCOR code (ver. 1.8.6).

Influence of Krypton Seeding on EU DEMO Operation with Lithium Divertor

DEMO reactor with liquid lithium divertor operation in presence of krypton seeding is analyzed. Integrated core-scrape-off layer-divertor code COREDIV is used. Sputtering model is revised and sputtering dependence on the divertor surface temperature fitted to the experimental data is added. It was found that Kr seeding is beneficial for DEMO operation. Plasma fuel dilution by lithium ions is reduced to the level of 1% for high level of Kr seeding and fusion power is comparable to the modelling results for argon or neon seeded DEMO with tungsten divertor. Simulations show that high confinement mode is possible in presence of Kr seeding (power across the separatrix stays above the L-H threshold).

Drift kinetic effects and local current drive induced modification of magnetic shear on sawtooth activity in EU DEMO

The sawtooth activity is investigated for an EU DEMO reference plasma, including kinetic effects from both thermal particles and fusion-born alphas. Kinetic effects are studied in conjunction with modification of the magnetic shear near the q= 1 surface (q is the safety factor) due to local current drive. Kinetic stabilization on the internal kink mode is found by both the non-perturbative MHD-kinetic hybrid code MARS-K (Liu et al 2008 Phys. Plasmas 15 112503) and the perturbative semi-analytic Porcelli model. The latter predicts full stabilization of the mode, when the local magnetic shear s 1 at the q= 1 surface is less than 1 (e.g. s 1 ∼ 0.6 as is the case for the target EU DEMO plasma without local current drive). By increasing the local magnetic shear with local current drive up to s 1 ∼ 2.8, one of the Porcelli sawtooth crash criteria can be satisfied, by accessing the ion-kinetic regime. Direct MARS-K non-perturbative eigenvalue computations, on the other hand, predict less kinetic stabilization of the internal kink and more robust triggering of the sawtooth crash in EU DEMO. The general trend of the predicted results remains similar though, between the MARS-K model and the Porcelli model.

3D transient CFD simulation of an in-vessel loss-of-coolant accident in the EU DEMO fusion reactor

Accidental transients in fusion reactors, such as the in-vessel loss-of-coolant accident (LOCA) considered here, are generally analyzed using system-level codes. However, because of their lumped nature, such tools cannot predict local pressure and temperature values, which are instead of interest to assess the integrity of the vacuum vessel (VV) structures. It is then fundamental to prove that the system-level tools’ predictions are at least conservative. In this work, we analyze the helium flow inside the VV following a LOCA in the helium-cooled blanket of the EU DEMO, using a 3D transient computational fluid-dynamics (CFD) model implemented in the commercial STAR-CCM+ code. In view of the large pressure ratio, a hypersonic flow regime (Ma > 5) develops, with the formation of shock fronts requiring a high time and space resolution. The model is applied to compute the evolution of the pressure distribution inside the VV and then to compare it with the information provided by the system-level GETTHEM model. The predictions by the two models of the intervention time of the VV pressure suppression system are compared, showing that the 0D model is conservative.

Design and analysis of the secondary circuit of the DEMO fusion power plant for the HCPB BB option without the energy storage system and with the auxiliary boiler

The EU DEMO (European DEMOnstration Fusion Power Plant) is being designed to produce fusion electricity at a level of several hundred MW, by about 2060. The Primary Heat Transfer System (PHTS) transfers heat from the fusion reactor heat sources, to the Power Conversion System (PCS) responsible for the generation of electricity. Four cooling concepts for the DEMO Breeding Blanket (BB) and the related PHTS are considered. In some variants the Intermediate Heat Transfer System (IHTS) including the Energy Storage System (ESS) has been added between the PHTS and PCS, in order to mitigate transient effects resulting from the pulsed plasma operation, but other options with a direct coupling of the PHTS with the PCS are also being investigated. In the present work a detailed GateCycle model of the steam/water PCS, for the option Helium Cooled Pebble Bed BB without the IHTS/ESS and with an auxiliary boiler is created and its operation at the nominal conditions (plasma pulse) and at thermal power reduced down to ∼50 % of the nominal value (dwell period) is studied. The proposed model of the PCS cycle is used to demonstrate potential possibility of stable operation of a EU DEMO plant without the IHTS/ESS during both pulse and dwell phases. It was shown, that the proposed PCS circuit can operate with the gross electrical power of 1120 MW and 547 MW during the pulse and dwell phase, respectively.

Effect of impurity radiation and helium particle confinement on tokamak–reactor plasma performance

In this paper two important matters for assessments the D–T plasma performance of future tokamak–reactors such as ITER and EU–DEMO projects are discussed. The first issue is the heat removal via the radiation of intrinsic and seeded impurities. The second is the helium particle confinement and ash removal. To study these issues, a simple numerical 0.5D-model is proposed. The impurity radiation in the tokamak high–temperature plasma is described by the coronal model. Coronal radiative cooling used rates are based on the tabulated data published in the recent works. Helium particle confinement and ash exhaust efficiency are determined by the variable parameter ρ* characterizing the ratio between the global helium particle confinement time and the energy confinement time τ E . Such approach allows simultaneously obtaining solutions of the particle and energy balance equations in steady–state conditions, taking account of the effect of impurity radiation loss and ash poisoning. Using known energy confinement scaling τ E and design parameters of the ITER and EU–DEMO projects the particle and energy balance equations have been solved. For fixed values of impurities concentration and ratios ρ*, iso–curves of constant fusion power were obtained. These curves are similar to the well–known Plasma OPeration CONtours (POPCON). The POPCON curves are often used to study the operating points of tokamak–reactors. It is shown that for the ITER parameters there is a noticeable margin in the ratio ρ*. For the design parameters of the EU–DEMO project, the margin for ratio ρ* is very small, which casts doubt on the feasibility of the project in its current form (at least with regard to the EU–DEMO steady–state project). Further research in this area is required. The methods and results obtained in this paper can be used for further conceptual analysis of tokamak–reactors operating space.

Industrial-scale manufacturing experience of titanium beryllide block for DEMO blanket application

Abstract For the first time, the TiBe12 blocks have been manufactured by powder metallurgy using industrial equipment at the Ulba Metallurgical Plant. Such blocks will be used for neutron multiplication in the helium-cooled breeding blanket for the EU DEMO fusion reactor. The vacuum annealing of a cold-pressed composite of Be and Ti powders at 600–1275 °C approved the conditions for TiBe12 synthesis. After hot isostatic pressing at 1150 °C, a single-phase TiBe12 structure was formed, but the sample fragmented into several pieces. These beryllide pieces were used to manufacture TiBe12 powder. Vacuum hot pressing of the obtained powder resulted in a solid titanium beryllide workpieces of O150 mm × 170 mm. Electrical discharge machining and waterjet cutting were used to machine the external and internal surfaces in accordance with the design requirements. The obtained block has a density of 98.3 % of the theoretical density of TiBe12. Microstructural analysis revealed a bimodal grain size distribution with grains of 5–10 μm and 20–40 μm. Electrical conductivity and thermal expansion coefficient of TiBe12 are similar to other beryllides. To mimic the pulsed operating conditions in DEMO, a sample of titanium beryllide was rapidly heated to 900 °C and cooled with a rate of 12 K/s. After 50 cycles, the sample had no cracks or severe surface oxidation.

Analysis of the influence of the different impurity seeding on the burn-up fraction and plasma confinement in the EU DEMO reactor

Abstract This work describes integrated numerical modelling applied to DEMO discharges with tungsten wall and divertor, using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multi-fluid transport in the SOL. The model is self-consistent with respect to both the effects of impurities on the α-power level and the interaction between seeded (Ar, Kr and Xe) and intrinsic impurities (tungsten, helium). This work is to analyse the influence of the impurity seeding on the burn-up fraction and plasma confinement in EU DEMO reactor. For the simulation with constant electron density at the separatrix, it is found that impurity seeding has a small influence on the burn-up fraction, which remains around 6 – 7%., but fuelling source is reduced from 2.7 to 2 × 1022 1/sec when moving from lowest to highest seeding level. The degradation of confinement for high Z seeding impurity, which is correlated to radiation in the core (due to seeded Kr and Xe) is observed. Better confinement for Ar at middle and high seeding level is observed in comparison to the case without seeding, because Ar radiation is small in the core (about 40-55 MW).

Analysis of the effects of primary heat transfer system isolation valves in case of in-vessel loss-of-coolant accidents in the EU DEMO

Abstract As DEMO is the first European device planned to produce electricity from fusion, the volume of its Primary Heat Transfer Systems (PHTS) will be consistently larger if compared to present or next-generation tokamaks such as ITER. The consequences of an in-vessel Loss-Of-Coolant Accident (LOCA) would then be more important, and within the EUROfusion Consortium different possible mitigation measures are being investigated. Among these, the introduction of Isolation Valves (IsoVs) on the main cooling loops of the Breeding Blanket is being considered, in view of the many benefits they would introduce, not only in case of accidents, but also e.g. during the maintenance of the in-vessel components. Fast-closing IsoVs on the PHTS would help in relaxing not only the requirements of the VV pressure suppression system (VVPSS) design, but also those related to the expansion volumes that shall accommodate the contaminated coolant discharged from the PHTS after a LOCA. In the present work, the GETTHEM code, the system-level thermal-hydraulic model developed for the EU DEMO at Politecnico di Torino, is used to assess the beneficial effects of the introduction of the IsoVs. The effects of the actuation time of the IsoVs and of their location are parametrically investigated, considering both water and helium as PHTS coolants, with particular reference to the reduction of the in-vessel space-averaged pressure and of the suppression system size.