Vacuum Vessel Pressure Research Articles

Performances of the ITER Pressure Suppression System during unstable steam condensation regimes

Abstract The paper deals with the qualification of the main safety system of the nuclear fusion reactor ITER: the Vacuum Vessel Pressure Suppression System (VVPSS). This safety system manages the Loss of Coolant Accident (LOCA), that could occur in the Vacuum Vessel (VV), avoiding a significant increase of the internal pressure that could breach the primary confinement barrier. Steam condensation occurs at sub-atmospheric pressure. This condition is original respect the usual applications of steam direct condensation in nuclear fission reactors. An extensive experimental research program, funded by ITER Organization, on reduced scale experimental rigs (scale 1/22 and 1/10) have been carried out at the University of Pisa. Unstable condensation regimes occurred during some tests at about 500 g/s of steam mass flow rate (10 % of the maximum steam mass flow rate foreseen for the relevant accidental scenarios). In these conditions, high intensity vibrations of the whole experimental rig occurred. These events can be classified as ‘Chugging’ or Condensation Induced Water Hammer (CIWH). 
The performed tests have been analysed elaborating the images of the video-cameras located in the pool and comparing the steam jet shape with the acceleration values recorded by an accelerometer mounted on the sparger structure.
These elaborations permitted to evaluate the dynamic of bubble collapse, the volume of the developed external bubbles, the collapse frequency as function of thermal-hydraulic parameters. 
Moreover, the paper emphasizes the beneficial effects of non-condensable gases on the unstable condensation regimes. In fact, the non-condensable gas inhibits the occurrences of Water Hammer inside the sparger and eliminates the thermal stratification inside the water pool increasing the turbulence.

Computational Thermal-Fluid Dynamics analyses of borated water distribution in the Vacuum Vessel of the Divertor Tokamak Test facility

This work presents the full set of 3D Computational Thermal-Fluid Dynamics (CFD) analyses of the Divertor Tokamak Test (DTT) Vacuum Vessel (VV). The VV D-shaped double shell structure, divided into 18 Regular and Special sectors, is actively maintained at the operating temperature of 60 °C by borated water in forced flow to counteract the thermal loads and the heat transfer with the surroundings. Due to manufacturing and integration constraints, the different sectors cannot share the same design: the water will flow in the free space between the two shells composing the VV but, given the complexity of the geometry, a careful hydraulic design is mandatory, to avoid local stagnation points which may cause either overheating or freezing. The different hydraulic paths are separately analysed with the Star-CCM+ software, with a SST k−ω turbulence closure, proving the effectiveness of their design (pressure drop, temperature constraints and stagnation/backflow). In addition, the borated water mass flow rate distribution among the different paths is assessed. To conclude, results from CFD analyses are exploited to approximate the hydraulic characteristic of each sector and to develop a system-level model of the full VV with the Modelica language. The overall VV pressure drop of Δp=2810 Pa, the outlet mixing temperature of Tout,mix=333.06 K and a mass flow rate distribution close to the homogeneous condition (≈2.22 kgs−1 in each sector) confirm that no relevant issues are found in the current VV design from the thermal-hydraulic point of view.

Preliminary thermal-hydraulic deterministic safety analysis of an in-vessel LOCA for the DTT facility

The DTT (Divertor Tokamak Test) facility is a new experimental tokamak under development by an Italian consortium in cooperation with several high-standard European laboratories. The ENEA division FSN-SICNUC is involved in the project for carrying out deterministic safety analysis of postulated accidents. In the first year of activity, it has been analyzed an In-Vessel LOCA (IVLOCA) scenario. The IVLOCA scenario selected is caused by a break in the divertor’s cassette cooling tubes with the consequent release of coolant inside the Vacuum Vessel (VV). The best estimate thermal-hydraulic system code TRACE, developed by USNRC, was selected to conduct the preliminary thermal-hydraulic analysis of this accident. DTT is currently under design, so not all the data are frozen. Therefore, based on some engineering assumptions and scaling considerations from similar facilities, the nodalization of DTT VV was developed using the three-dimensional TRACE component “VESSEL”. Then, a sensitivity analysis was carried out to simulate the break and the consequent water injection in the VV. This was done to compare the system behavior and test different nodalization approaches. Subsequently, the data of the DTT divertor cooling system were used to run additional simulations. The results allow to compare different nodalization approaches and to have a preliminary estimate of the pressure peak and temperature behavior in the VV for an IVLOCA. Finally, a first uncertainty analysis was carried out using the DAKOTA toolkit, coupled with TRACE code in SNAP. Two uncertain input parameters were selected: the rupture area of a cooling divertor tube and the temperature of the divertor coolant. The uncertainty analysis allows having a wider spectrum of system behavior and a preliminary insight on the dispersion of the VV pressure, selected as figure of merit. This paper aims to show the results of this preliminary analysis, characterizing the phenomena that occurred during the selected transient.

Tritium transport in the vacuum vessel pressure suppression system for helium cooled pebble bed

In the frame of the safety studies for the EU-DEMO reactor, attention is paid to the hydrogen concentration in the vacuum vessel and connected volumes since it would lead to a possible hazard of releasing tritium and activated dust. The risk of explosion cannot be excluded a priori if H2 stockpiles. For this reason, in both water (WCLL) and helium (HCPB) cooled breeding blanket concepts of EU-DEMO, the problem is under investigation with a cross-reference between the available technologies in fission (such as the Passive Autocatalytic Recombiners – PAR) and fusion application. In particular, the recent analyses pointed out the implementation of the PARs into the Vacuum Vessel Pressure Suppression System or linked systems.This paper evaluates the Hydrogen behavior (main mobilized tritium source term) for the Helium-Cooled Pebble Bed (HCPB) VVPSS concept. The analyses preliminary investigate the stratification of the hydrogen mass inventory inside the PSS. In particular, a MELCOR 1.8.6 model of the PSS, based on past activities aimed at dust transport and thermohydraulic analyses, is adopted. The paper also introduces the applicability of PAR technology in the operation range of fusion devices, analyzing the problem of the recombination rate due to the dilution of Hydrogen after a Helium blowdown.

Reverse flow in a pressure suppression system due to condensation instabilities

The Vacuum Vessel Pressure Suppression System (VVPSS) is designed to protect the ITER reactor (Vacuum Vessel, VV) from an overpressurization accident. The VVPSS is mainly composed of four Vapour Suppression Tanks (VSTs) partially filled with water inside which venting pipes (spargers) are immersed to direct the steam in the subcooled pool to condense it. Since the maximum pressure allowed inside the VV is only 0.15 MPa, a sub-atmospheric pressure must be maintained inside the VSTs. During a series of experimental campaigns funded by ITER Organization to study and validate the correct operability of the VVPSS and conducted in a large-scale plant built at the B. Guerrini laboratory of the University of Pisa, pressure instability and reverse flow through the sparger holes have been observed. The analysis of acceleration signals recorded during this pressure inversion, showed large amplitude oscillations at low frequency due to the Condensation Induced Water Hammer (CIWH).

Passive Hydrogen Recombination during a Beyond Design Basis Accident in a Fusion DEMO Plant

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO.

Effect of different system parameters on the design of the EU DEMO Vacuum Vessel Pressure Suppression System

The Vacuum Vessel Pressure Suppression System (VVPSS) is the safety system devoted to the mitigation of in-vessel Loss-Of-Coolant Accidents, which are considered among the Design-Basis Accidents for tokamak fusion reactors. The EUROfusion Consortium is performing the design of this system, as part of the conceptual design phase of the EU DEMO fusion reactor, including in the process its integration in the plant.This work reports the first parametric calculations carried out to evaluate the VVPSS response to a postulated double-ended guillotine rupture of a Breeding Blanket coolant feeder. The aim of the study is to characterize the role that key parameters play on the multiple systems involved in such an accidental event. Assessments scanned a broad range of design variables, some closely related to the suppression system itself, and others dependent on Primary Heat Transfer System features. The parameter space is mapped identifying the regions where the accident consequences are fully mitigated, in terms of peak pressure in the Vacuum Vessel (VV) and equilibrium pressure of both VV and VVPSS, providing a design tool to rapidly adapt the VVPSS to the evolving plant design.

Dust produced by plasma off normal event in the vacuum vessel: Experimental analyses of the deposition inside the pressure suppression tank during a LOCA

This paper deals with experimental tests simulating the dust deposition, produced in the Vacuum Vessel of the nuclear fusion reactor ITER, and carried by steam and gases inside the Pressure Suppression Tanks during a Loss of Coolant Accident (LOCA). The dust, mainly tungsten and beryllium, is produced by the erosion of the divertor or of the first wall due to plasma off-normal events, like plasma disruptions or vertical displacement events. The activity is performed in the frame of a research program performed at the University of Pisa, funded by the ITER Organization, for studying the behaviour of the Vacuum Vessel Pressure Suppression System (VVPSS). A reduced scale of VVPSS was built at the University of Pisa (1/32 geometrical scale). The experimental rig is composed by an electric steam generator and steam and air mass flow rate control lines. Steam with (or without) air is sent in a condensation container, partially filled with water. These fluids entrain a mass of dust measured by a dust mini-dosing system. The air and the non-condensed steam flow in the atmosphere through a demister and HEPA filters. A tests campaign was carried out injecting a simulant of the actual dust (aluminium oxide) in a flux of steam and air into the condensation tank. The tests were performed at different values of steam, air and dust mass flow rate. During the test, the dust concentration in the water was measured versus time, in order to determine the settling time. The dust deposition on the surfaces of the condensation tank was measured as well as the decontamination efficiency of the tank. The results demonstrated that about 99 % of dust is deposited on the tank surfaces and only a quantity less than 1 % is carried by the water during the tank discharge.

Parametric analysis of influence factors to the in-vacuum vessel LOCA for CFETR based on RELAP5

The water-cooled blanket and water-cooled divertor located inside the vacuum vessel (VV) are two crucial components of the Chinese Fusion Engineering Testing Reactor (CFETR). The flash vaporization from high temperature and high pressure water pressurizes the VV when in-VV Loss of Coolant Accident (in-VV LOCA) occurs. As the first confinement barrier of the CFETR, the integrity of the VV is required to be guaranteed to contain radioactive materials. Parametric analysis of the in-VV LOCA is carried out to analyze the key influence factors on the pressure increasing rate and leaked water mass so as to provide basis for safety facility designs of the CFETR. The water-cooled components, VV and vacuum vessel pressure suppression system are modeled using RELAP5. The pressure increasing rate and leaked water mass are analyzed under circumstances with different discharge pressures, VV initial pressures and the breach areas. Moreover, the pressure increasing rate increases with the discharge pressure, VV initial pressure and breach area increased. The leaked water mass and its variation both increase with breach area increased.

Highly underexpanded helium jet and its effect on dust resuspension under in-vessel loss of coolant accident of fusion reactor

During an in-vessel loss of coolant accident (LOCA) of a helium-cooled blanket in a fusion reactor, the high-pressure helium gas ingress into the vacuum vessel (VV) will produce highly underexpanded jet, resulting in the resuspension and migration of the radioactive dust deposited in the VV. In this study, the flow field in the VV and the influence of friction velocity were simulated and analyzed by using ANSYS Fluent code. In addition, the differences between in-vessel LOCA and loss of vacuum accident (LOVA) were compared. The results showed that the pressure of the VV reaches the pressure relief limit of the VV pressure suppression system rapidly after a LOCA. With the development of a LOCA, the Mach disk continued to move towards the equatorial break of the VV, its shape continued to shrink and the number of structures gradually increased. Compared with the LOVA, the LOCA had a higher friction velocity, resulting in a greater probability of dust resuspension and radioactive risk.

Effect of non-condensable gas in steam condensation at sub-atmospheric pressure condition

In the International Thermonuclear Experimental Reactor (ITER), a postulated Loss Of Coolant Accident (LOCA) in the Vacuum Vessel (VV) has to be managed with a pressure suppression system working at sub-atmospheric pressure. The operating conditions considerably differ from those experienced in the fission nuclear power plants such as BWR, since the ITER Tokamak works at very low pressure conditions and can withstand a maximum pressure of 0.15 MPa. For this reason, the pressure value must not exceed 10 kPa for a water temperature of 30 °C inside the Vapour Suppression Tanks (VSTs) that are the fundamental components of the Vacuum Vessel Pressure Suppression System (VVPSS). During a LOCA some non-condensable gases (mainly hydrogen and oxygen gases due to the water radiolysis or thermolysis) may be mixed in the steam and this could impair the condensation efficiency. In order to investigate the effects of non-condensable gas on DCC, we conducted a research program funded by ITER Organization at the laboratory of the University of Pisa: we designed and built a small-scale experimental rig to study the steam Direct Contact Condensation (DCC) with the presence of non-condensable gas and simulate the behaviour of a VST. Since DCC can occur with different characteristics, we ran 12 closed mode tests exploring all condensation regimes injecting a certain mass of air with the steam discharged in the subcooled water. The tests started at the saturation pressures corresponding to water temperatures ranging from 40 °C to 80 °C and ended when the free space volume reached the atmospheric pressure. From the analysis of the data acquired during the tests we observed that the condensation efficiency remained higher than 95%. We observed that despite this, the presence of a certain quantity of non-condensable gas has negative aspects on condensation: the condensation regime never reaches stability (the regimes quickly shift towards instability). Furthermore, the presence of air triggers a turbulence of the flow which interferes with the transfer of heat from the steam to the water. Not only the introduction of air into the flow increases linearly the pressure above the water head but its high temperature further contributes to the pressure increase. The air mixed with the flow also forms eddies that can trap the steam and transport it out of the water, preventing it from condensing. Apart from condensation, the most noticeable problem is the rapid increase in pressure inside the condensation tank.

Experimental Analysis of Steam Condensation Heat Transfer Coefficient in Water Pool at Sub-Atmospheric Pressure

The experimental analysis of steam direct contact condensation in a water pool at sub-atmospheric condition was carried out in relevant configurations for the International Thermonuclear Experimental Reactor (ITER) Vacuum Vessel Pressure Suppression System, during a postulated Ingress of Coolant Event category IV accidental scenario. This transient accident was experimentally simulated in a reduced scale (1:22) facility at University of Pisa, adopting a defined scaling law. The steam jet plumes were video recorded and image analysis was performed at stable and interfacial oscillation condensation regimes, providing steam plume average length and interfacial area. The average heat transfer coefficient was therefore computed and correlated to the condensation driving potential. Also, the steam jet length divided by orifice diameter (L/D) was characterised as function of driving potential. Empirical correlations available in literature at atmospheric conditions were modified for evaluating vacuum conditions peculiarities. The remarkable higher dependence of heat transfer coefficient (HTC) to driving potential parameter, at vacuum conditions, was highlighted by the driving potential exponent equal to 1.2 instead of 0.04 at ambient pressure. The obtained L/D and HTCcorrelations are able to predict with ± 15% and ± 30% of error the corresponding experimental data, respectively.

Deposition of ITER Vacuum Vessel Dust Inside the Pressure Suppression System During a Loss of Coolant Accident: Experimental and Numerical Analyses

Abstract This paper deals with an experimental and numerical analysis of the deposition of International thermonuclear experimental reactor (ITER) dust simulant inside a reduced scale vacuum vessel pressure suppression system (VVPSS) of ITER. This research, funded by the ITER organization, aims to analyze the dust deposition in a water container in a relevant configuration for the ITER VVPSS, the dust removal by means of robotized apparatuses and their decontamination efficiency. The experimental rig, built at the department of civil and industrial engineering (DICI) of the University of Pisa (Italy), is described and the results of a preliminary experimental test are illustrated. It was measured that two and 82% of the overall injected dust mass was strongly and lightly bound to the water container surfaces, respectively, and 16% were not deposited in the water tank. Furthermore, three numerical analyses were carried out implementing a model of the experimental rig in the Enel code for analysis of radionuclide transport (ECART) to determine the relevance of different parameters on the deposition, resuspension and removal of dust. The numerical simulations allowed to specify dust mass deposition on the different rig components, revealing a strong dust retention (about 66%) in the first part of the injection piping in case of coarse granulometry. Finest lognormal dust distribution was instead able to reach the water container (about 90%). Moreover, the numerical simulations permitted to define more precisely the test matrix and to analyze the experimental results.

Integrated design of breeding blanket and ancillary systems related to the use of helium or water as a coolant and impact on the overall plant design

Currently, for the EU DEMO, two Breeding Blankets (BBs) have been selected as potential candidates for the integration in the reactor. They are the Water Cooled Lithium Lead and the Helium Cooled Pebble Bed BB concepts. The two BB variants together with the associated ancillary systems drive the design of the overall plant. Therefore, a holistic investigation of integration issues derived by the BB and the installation of its ancillary systems has been performed. The issues related to the water activation due to the 16N and 17N isotopes and the impact on the primary heat transfer systems have been investigated providing guidelines and dedicated solution for the integration of safety devices as isolation valves. The tritium retention and the permeation rates through the blanket and its ancillary systems have been also assessed taking into account different operating points both for the BB and ancillaries and comparing, when possible, the releases with the operating and safety limits. Moreover, the issues related to the tritium start-up inventory as well as the uncertainties on the Tritium Breeding Ratio (TBR) due to the integration of the auxiliary systems within the Vacuum Vessel have been also studied. Finally, the impact of the BB concepts on the safety systems like the Vacuum Vessel Pressure Suppression System is described with a particular focus on the different measures that should be implemented according to the considered concept. All these aspects are then taken into account to drive future developments during the Concept Design Phase.

Concept of operation of the remote handling system for the ITER vacuum vessel pressure suppression system

A concept of the remote handling (RH) tools to maintain the components of the ITER Vacuum Vessel Pressure Suppression System (VVPSS) is developing. These RH tools are designed to remove and replace the rupture disk assembly and the bleed line valve assembly, to clean the relief lines, and to perform any associated operations to maintain them. The RH tools work together with the Neutral Beam Remote Handling System (NBRHS). The main functions of the RH tools are disconnection and connection of the bolted flanges, keeping the static confinement during the RH operations, limiting the spread of contamination, and vacuum or water jet cleaning of the contamination inside the relief line. The RH interfaces in the components are defined during the design process of the RH tools in collaboration with the VVPSS component designers. An RH sequence study is being carried out to demonstrate the feasibility of the maintenance operations within the constrained NB cell environment, and the capability of the NBRHS.

Hydrogen explosion mitigation in DEMO vacuum vessel pressure suppression system using passive recombiners

An important issue for the Fusion Reactors is the hydrogen explosion hazard assessment. For this reason, in the Work Package of Safety Analyses and Environment (WPSAE) of the EUROFusion consortium, a task was established to identify the potential for hydrogen production in DEMO vessel in accident situations and to investigate the possible solutions which can prevent the risk of hydrogen and dust explosion. The hydrogen in the Vacuum Vessel (VV) can lead to combustion progressing in deflagration and detonation. Besides, the tungsten dust could enhance the effects of H2 reaction as demonstrated by experiments, however, in this first set of accidents, only the H2 generated by the chemical reaction is accounted for.For these reasons, the solutions to avoid or limit the H2 production and/or to control safely the risk of H2 accumulation and explosion need to be screened starting from the solutions adopted in the fission technology, and later on, adapted and assessed in the Pressure Suppression System (PSS) foreseen in DEMO. One promising idea seems to install passive catalytic recombiners in the PSS.The aim of this paper is firstly to set a range of operating parameters for the PAR functioning and intervention suitable in DEMO PSS configuration. Secondly to verify if the present criteria and parameters work efficiently in mitigating the H2 accumulation a PSS in accidental conditions. To test the theoretical effectiveness of the PAR intervention, two reference accidents have been analyzed: an in-VV Loss Of Coolant Accident (LOCA) in the frame of the Design Basis Accident (DBA) and a Loss Of Flow Accident (LOFA) without active plasma shutdown in the context of the Beyond Design Basis Accidents (BDBA).The results seem promising if the performances of the PAR will be confirmed in sub atmospheric and saturated conditions.

Large scale experimental facility for performance assessment of the vacuum vessel pressure suppression system of ITER

The nuclear fusion reactor ITER (International Thermonuclear Experimental Reactor) foresees a Pressure Suppression System (PSS) in order to manage a Loss of Coolant Accident (LOCA) in the Vacuum Vessel (VV). A LOCA scenario is postulated to occur in the cooling systems of plasma facing components. The Vacuum Vessel Pressure Suppression System (VVPSS) has a key safety function because a large internal pressure in the VV could lead to a breach of the primary confinement barrier.Steam condensation would occur in ITER at sub-atmospheric pressure, differently than the applications in nuclear fission reactors. Steam direct contact condensation (DCC) in sub-atmospheric conditions have never been implemented before, therefore an experimental assessment of the VVPSS is necessary to simulate this phenomenon in ITER's thermal-hydraulic conditions.Under the financial support of ITER Organization, an experimental program was performed at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa in the past 3 years, using a small-scale experimental rig (1/22 scale factor).In order to assess the scale laws elaborated by means of the experimental results obtained in this small scale facility, a Large Scale facility was built at University of Pisa. This facility is almost a full scale mock-up of ITER's VVPSS from the geometric point of view (1/1.09 scale factor) and it is 1/10 scale factor from the steam mass flow rate point of view.This paper illustrates the results of the first test campaign. A 1/10 scale sparger system SPS-A (having a diameter equal to DN125 and 100 holes) was used with a maximum steam mass flow rate of 0.5 kg/s. This experimental configuration represents real operating conditions on the basis of the results of elaborated similitude analysis. The preliminary analysis of the experimental results confirms the agreement in terms of condensation regimes between the two reduced scale (1/22 and 1/10 scale) as foreseen by the elaborated scale laws.

Direct steam condensation at sub-atmospheric pressure: experimental tests and similitude analysis

This paper deals with experimental and theoretical analyses of the steam direct condensation at sub-atmospheric pressure. These operation conditions occur in the Vacuum Vessel Pressure Suppression System (VVPSS) of ITER Nuclear Fusion Reactor. This safety system permits to manage the Ingress of Coolant Event (ICE cat. IV) accident condition, postulated to occur in the ITER Vacuum Vessel (VV) as a consequence of the rupture of the shielding blanket cooling piping or the rupture of the divertor cassette piping.Experimental tests of steam direct condensation at sub-atmospheric pressures in a 1:22 scale facility were performed at the laboratory Guerrini of DICI-University of Pisa. Presently a full-scale facility is being building at DICI. This paper illustrates the similitude analysis elaborated to scale up the experimental results obtained in the reduced scale facility. The main variables which influence the steam direct condensation were analyzed and the relative scale laws determined. CFD numerical simulations of direct steam condensation in the actual geometry and in a vessel of equal volume permitted to verify some of the assumptions on which the scale laws are based. Scaling studies were first performed in order to calculate the transient of steam mass flow rate occurring during the ICE IV event in the scale apparatus, starting from the thermal hydraulic studies performed at ITER and successfully compared with the experimental results carried out in scaled apparatus. This confirmed the suitability of scale laws and in the same time the capability of the VVPSS to condense the injected steam at sub-atmospheric pressure, matching the safety goal to reduce the system pressurization.

DEMO Divertor preliminary safety assessment

European DEMOnstration plant is currently finalizing a pre-conceptual design phase. A carefully selected set of safety analyses are contextually being performed on DEMO systems in order to assess its integrated performance and capability to achieve expected targets while keeping in a safe operation domain. A key in-vessel system is the Divertor, in charge of exhausting a major part of the plasma ions thermal power in a region far from plasma core to control plasma pollution. To accomplish this mission a modular design approach was adopted based on two main sub-systems: the divertor cassette and plasma facing units (PFUs). Each cassette supports two vertical plasma facing units, one inboard and one outboard, intercepting the main part of the particle loads from the plasma. An assembly of 3 of such divertor cassettes and related PFUs is then foreseen for each of DEMO 16 sectors, hence resulting into 48 cassettes. Two independent heat transfer systems for cassettes and plasma facing units respectively provide heat sink function and connect the Divertor to DEMO Balance of Plant.An overview of the preliminary safety analyses performed on such system is here proposed. At first, a safety-perspective overview of current design is proposed. In particular, moving from the divertor system functions, specific component failure modes and related possible consequences, a set of relevant postulated initiating events are identified.Then, aiming at investigating whether accident consequences remain within accepted safe domain, two accidents from this set were selected to be analysed deterministically by means of MELCOR code. In particular, an in-vessel Loss of Coolant Accident initiated by a large rupture of the divertor PFU cooling pipes and a Loss of Flow Accident (LOFA) initiated by heat transfer system (HTS) pump seizure have been analysed. The nodalization includes divertor PFU and related HTS among other in-vessel components and safety devices (e.g., Vacuum Vessel Pressure Suppression System) devoted to Vacuum Vessel pressurization transient control or Fast Plasma Shutdown System. Parametric analyses were also performed to investigate possible impact of the design solutions (e.g., implementation of isolation valves in Divertor loop) on the accident consequences.

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).