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Activated Corrosion Products (ACPs) formation and deposition pose a critical safety issue for nuclear fusion reactors. The working fluid transports the ACPs towards regions accessible to worker personnel. Predicting ACPs formation deposition and transport is fundamental for source term identification, reduction of radiation exposure assessment, maintenance plan definition, design optimization, and waste management. The code OSCAR-Fusion has been developed by the CEA (France) to evaluate the ACPs generation and transport in closed water-cooled loops for fusion application. This work aims at assessing the impact of water chemistry on the transport, precipitation, and deposition of corrosion products for the EU-DEMO divertor Plasma Facing Unit Primary Heat Transfer System. Sensitivity analyses and uncertainty quantification are needed due to the multi-physics phenomena involved in ACPs formation and transport. The OSCAR-Fusion/RAVEN code coupling developed by the Sapienza University of Rome and ENEA has been used. This work presents the perturbation results of different parameters chosen for a closed water-cooled loop considering a continuous scenario of 1888 days. The aim of this work is to assess the variation of build-up of ACPs, perturbing the alkalizing agent concentration into the coolant and the corrosion and release rates of different materials.

After decades of operation in nuclear power plants, Once-Through Steam Generators (OTSGs) were recently proposed for nuclear fusion applications. In particular, they are supposed to be installed in the primary cooling systems of the European Union Demonstration fusion power plant (EU-DEMO). One of the key reactor components is the Breeding blanket (BB). Among the BB concepts that are currently under study, Water-Cooled Lithium-Lead (WCLL) option was considered for this work. The WCLL blanket is divided in two main subsystems, the breeder zone (BZ) and the first wall (FW), each one provided with an independent cooling circuit, named Primary Heat Transfer System (PHTS). Thermal power removed from BB by BZ and FW PHTS is driven to the Power Conversion System (PCS) to be converted into electricity. The thermal coupling is ensured by two OTSGs per system. At the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of Sapienza University of Rome, a simulation activity was carried out to understand the component thermal-hydraulic behavior during DEMO normal operations. For calculation purposes, a full model of the steam generator was prepared by using a modified version of RELAP5/MOD3.3 system code. The computational activity performed allows to preliminary characterize the OTSG thermal-hydraulic performances during both pulse and dwell phases.

The breeding blanket is one of the fundamental components of a nuclear fusion reactor and is responsible for the fuel production, generating tritium through neutronic capture reaction between lithium and neutrons. Lithium is a liquid PbLi alloy and the helium formed as reaction by-product can coalesce into bubbles, generating a two-phase mixture with a high-density ratio (ηρ∼O5). These bubbles can accumulate and stagnate within the blanket channels with potentially harmful consequences. In this work, the interIsoFoam solver of OpenFOAM v2012 is used to simulate bubble motion for a two-phase mixture representative of the He-PbLi system to test its potential for future developments in the field of fusion. In a first phase, several traditional benchmarks were carried out, both 2D and 3D, and considering the two variants of the VOF method implemented in the solver, isoAdvector and plicRDF. Subsequently, He bubbles of different diameters rising in liquid PbLi (ηρ=1.2×105) were analysed to investigate different regimes. For a Eötvös number (Eo) greater than 10, it was possible to recreate the axisymmetric, skirted, oscillatory regimes and the peripheral and central breakup regimes. For Eo < 10, non-physical deformations of the interface are observed, probably generated by spurious velocities that have a greater impact on the solution for very small bubbles and rising velocities.

Magnetohydrodynamic (MHD) phenomena, due to the interaction between a magnetic field and a moving electro-conductive fluid, are crucial for the design of magnetic-confinement fusion reactors and, specifically, for the design of the breeding blanket concepts that adopt liquid metals (LMs) as working fluids. Computational tools are employed to lead fusion-relevant physical analysis, but a dedicated MHD code able to simulate all the phenomena involved in a blanket is still not available and there is a dearth of systems code featuring MHD modelling capabilities. In this paper, models to predict both 2D and 3D MHD pressure drop, derived by experimental and numerical works, have been implemented in the thermal-hydraulic system code RELAP5/MOD3.3 (RELAP5). The verification and validation procedure of the MHD module involves the comparison of the results obtained by the code with those of direct numerical simulation tools and data obtained by experimental works. As relevant examples, RELAP5 is used to recreate the results obtained by the analysis of two test blanket modules: Lithium Lead Ceramic Breeder and Helium-Cooled Lithium Lead. The novel MHD subroutines are proven reliable in the prediction of the pressure drop for both simple and complex geometries related to LM circuits at high magnetic field intensity (error range ±10%).

In the Water-Cooled Lithium Lead (WCLL) blanket, a critical problem faced by the design is to ensure that the breeding zone (BZ) is properly cooled to avoid the loss of mechanical properties in the structural materials. CFD simulations are performed using ANSYS CFX to assess the cooling system performances accounting for the magnetic field effect in the sub-channel closest to the first wall (FW). Here, intense buoyancy forces (Gr ≈ 1010) interact with the pressure-driven flow (Re ≈ 103) in a MHD mixed convection regime. A constant magnetic field, parallel to the toroidal direction, is assumed with Ha = 8550. The walls bounding the channel and the water pipes are modeled as perfectly conducting. The magnetic field is found to dampen the velocity fluctuations triggered by the buoyancy forces and the flow is similar to a forced convection regime. The PbLi heat transfer coefficient is reduced to one-third of its ordinary hydrodynamic value and, consequently, hot-spots close to the FW are observed, where TMax ≈ 1000 K. Optimization strategies for the BZ cooling system layout are proposed and implemented in the CFD model, thus fulfilling the design criterion.

The best estimate plus uncertainty (BEPU) method plays a key role in the development of the innovative Generation IV nuclear reactors, for the improvement of knowledge and the good evaluation of the safety margins for new phenomena. The aim of this paper is to validate an uncertainty quantification (UQ) approach using RAVEN code. RAVEN, developed at the Idaho National Laboratory, is a multipurpose probabilistic and uncertainty quantification framework, capable to communicate with any system code, implemented with an integrated validation methodology involving several different metrics. In this activity, a coupled calculation RELAP5-3D/RAVEN has been performed to assess the validity of the embedded UQ approach. The simulations have concerned two tests conducted on NACIE (NAtural CIrculation Experiment) facility, a non-nuclear loop-type system using Lead-Bismuth Eutectic (LBE) as coolant, realized at the ENEA Brasimone Research Centre (Italy) to support the development of the GEN-IV reactors. The experimental tests are aimed to investigate the phenomena related to the natural and gas enhanced circulation flow regimes of heavy liquid metals (HLM), and to test and validate the main components in a LBE environment. A numerical 1D model of NACIE facility has been realized and the post-test analysis has been carried out using RELAP5-3D© ver. 4.3.4. Based on these results, a statistical analysis has been performed using RAVEN computer code, investigating the capability of three probabilistic comparison metrics fully integrated in the code.

The design of the Chilled Water System (CHWS-1) of ITER (Nuclear Facility INB-174) considers two independent and segregated subsystems, CHWS-1 A & B, which provide cooling water to safety-relevant components in Tokamak and Hot Cell Complexes. CHWS-1 is a Safety Important System needed for the proper functioning of Protection/Safety Important Activities & Components and therefore Single Failure Criterion is included in its design, while physical separation between its trains is maintained as much as possible. However, both CHWS-1 trains are located in the Tokamak Complex Drain Tank Room (DTR) and thus, measures are taken in order to protect them from common causes of failure (i.e. Fire or High Energy Line Break). A failure of a given train in the DTR is investigated in order to analyse the consequences and take required additional measures so as to guarantee full safety cooling to CHWS-1 components outside DTR. A RELAP5 model of CHWS-1B is developed and a pipe break inside the DTR modelled, to perform thermal hydraulic transient analyses of this event. From the results, the transient evolution of the main thermal hydraulic parameters is determined and the required times of intervention evaluated in order to restore the cooling function of the affected CHWS-1 train for its remaining components. Thus, an accident management strategy is included in the analysis effort as well as a definition of CHWS-1 instrumentation and control fulfilling a relevant role in the event detection and mitigation. This paper summarizes the results of this activity.

A MELCOR model of a boiling water reactor has been developed by Sapienza University of Rome in the framework of the MUSA project aiming at identifying and quantifying uncertainty sources in severe accidents analyses. To develop the model, the Fukushima Unit 3 boiling water reactor has been taken as reference, and a preliminary sensitivity analysis with RAVEN software coupled with MELCOR has been performed to quantify the influence of core degradation parameters on selected key figures-of-merit, such as pressure vessel and containment pressures, core liquid level, hydrogen generation, lower head breach time and source term. Results of base case satisfactorily predict main pressure and liquid level data measurement from TEPCO. As well as, results from the uncertainty analysis envelope the majority of thermohydraulic TEPCO reported measured data. Furthermore, median of calculated core status is in likely agreement with recent TEPCO containment inspections and muon measurements, with about 25% of fuel rods remained intact in RPV and 65% of core masses ejected to the pedestal.

Liquid metals offer unique properties and their use in a nuclear fusion reactor, both as confined flows and free-surface flow, is widely studied in the fusion community. The interaction between thi...

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).

Permeator Against Vacuum (PAV) is one of the technologies proposed for the Tritium Extraction and Removal System (TERS) of the Water-Cooled Lithium Lead Breeding Blanket (WCLL BB). The paper presents the activity aimed at the engineering design of a PAV mock-up with a niobium membrane, in order to later assemble and qualify it. Experience gained in the engineering design of the mock-up, the heating system, the instrumentation, and the vacuum line is illustrated. This experience will be useful for the preliminary design, the manufacturing and the operation of the PAV with niobium membrane for DEMO. Niobium was selected as membrane material of this mock-up because of its high permeability and for its lower cost compared to vanadium, the other candidate material for membranes. Besides, niobium has a lower tendency to oxidation than vanadium. Oxidation would reduce the hydrogen isotopes permeation flux. In this paper, the solution adopted to manufacture the PAV mock-up, a complex component with niobium and P22 parts, is illustrated. The Nb/P22 welding issues are also presented, in particular related to the compatibility of the welded joints with LiPb. In the chosen design, the LiPb flows with two passages in 16 (8 + 8) niobium “U” shaped pipes installed in a vacuum chamber and welded to a P22 plate. The U-pipes configuration was selected to minimize the welding area, the volume of the component and the membrane thickness while trying to preserve the highest possible extraction efficiency.

The Breeding Blanket (BB) is one of the key components of the European Demonstration (EU-DEMO) fusion reactor. Its main subsystems, the Breeder Zone (BZ) and the First Wall (FW), are cooled by two independent cooling circuits, called Primary Heat Transfer Systems (PHTS). Evaluating the BB PHTS performances in anticipated transient and accident conditions is a relevant issue for the design of these cooling systems. Within the framework of the EUROfusion Work Package Breeding Blanket, it was performed a thermal-hydraulic analysis of the PHTS during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”, by using Reactor Excursion Leak Analysis Program (RELAP5) Mod3.3. The BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat Exchangers were included in the study. Selected transients consist in partial and complete Loss of Flow Accident (LOFA) involving either the BZ or the FW PHTS Main Coolant Pumps (MCPs). The influence of the loss of off-site power, combined with the accident occurrence, was also investigated. The transient analysis was performed with the aim of design improvement. The current practice of a standard Pressurized Water Reactor (PWR) was adopted to propose and study actuation logics related to each accidental scenario. The appropriateness of the current PHTS design was demonstrated by simulation outcomes.

In the framework of the HORIZON2020 SESAME European project, an experimental campaign has been carried out on the large Lead-Bismuth Eutectic pool integral effect CIRCE facility at ENEA Brasimone Research Center, implementing the HERO test section. This test section is aimed at supporting the development of the ALFRED design. Within the SESAME Project, three protected loss of flow accident tests have been designed and executed, identified as SE-Test1, SE-Test2, and SE-Test3. The third test (SE-Test3) has been selected by the project participants for a benchmark activity involving system thermal-hydraulic codes and coupled system thermalhydraulic/ CFD. This activity has been divided into two phases: a blind phase, and a post-test phase. The present paper illustrates the results achieved during the post-test phase of the CIRCE-HERO benchmark activity. Four participants simulated the experimental test. Three (ENEA, SCKCEN and UNIROMA1) used different versions of RELAP5 code, i.e. RELAP5-3D and RELAP5Mod3.3, and one (NRG) used a coupled approach based on an in-house system thermal-hydraulic code (SPECTRA) and the commercial code ANSYS CFX. The benchmark activity hereafter presented is a relevant exercise for evaluating and comparing the predictive capabilities of system thermal-hydraulic codes, CFD codes and coupled technics, in relation to phenomena occurring during the transition between forced and natural circulation (e.g. loss of flow) in a heavy liquid metals Generation IV system. The results of the numerical exercise showed an adequate capability of the codes to reproduce the relevant phenomena involved during the experiment. The SYS-TH codes have been more accurate than SYS-TH/CFD in predicting the trend of the main parameters during the transient, reproducing them with quite satisfactory accuracy. SYS-TH/CFD simulation, instead, provided a more satisfactory representation of the pool thermal stratification. The results highlighted that the planning and the execution of experiments fully devoted for the code V&V process is needed for the further development of such numerical tools, in particular for SYS-TH/CFD coupled approaches, whose use is more recent respect to SYS-TH codes.

The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium production, shield the Vacuum Vessel and remove the heat generated by plasma thermal radiation and nuclear reactions. It relies on PbLi eutectic alloy adopted as breeder and neutron multiplier and refrigerate by subcooled pressurized water. The last function is fulfilled by two independent cooling systems: First Wall (FW) that faces the plasma heat flux and the Breeding Zone (BZ) that removes the deposited power of neutron and photon interaction. Several layouts of WCLL BB system have been investigated in the last years to identify a configuration that guarantees Eurofer temperature below the limit (550 °C) and good thermal-hydraulic performances. This research activity focuses on the FW cooling system based on the WCLL BB 2018 design of DEMO 2017 baseline, investigating the cooling performances in order to optimize the FW design, reducing the amount of water that affects the Tritium Breeding Ratio (TBR) and improves the efficiency of the Primary Heat Transfer System (PHTS), verifying the reliability to deliver coolant at adequate design temperature (328 °C) to the PHTS. Different solutions have been considered and analyzed focusing on three specific positions along the poloidal direction: the equatorial cells in the Central Outboard Segment (COB) and Inboard Segment (IB), where there is the maximum deposited power and low Heat Flux (HF), and the apical cell of the IB, impacted by the highest HF but with low deposited power. For each FW design, several thermal-hydraulic steady-state analyses have been performed, as well as sensitivity analyses to evaluate the response of the systems under different cooling configuration, changing the position and the number of channels. Furthermore, a transient analysis reproducing the Dwell-Pulse phase of the DEMO fusion reactor with the optimized configuration of the Outboard FW layout has been performed to study the behavior of the FW cooling system subjected to the pulsed operation. The research activity aims at laying the basis for the finalization of the WCLL BB design, pointing out relevant thermal-hydraulic aspects. The analyses have been carried out using a CFD approach, thus a 3D finite volume model of each configuration has been developed, adopting the commercial ANSYS CFX code. The goal of the study is to compare the different FW cooling system layouts, identifying and discussing advantages and key issues from the thermal-hydraulic point of view. The results show that the coolant system of FW can safely remove the high plasma HF and nuclear heat deposition with a reduction of water channels, delivering coolant to the FW PHTS heat exchanger at the design temperature.

The development of a lead-cooled fast reactor (LFR) demonstrator was proposed, mainly in EU, to investigate the feasibility of an industrial size ELFR (European Lead-cooled Fast Reactor). The demonstrator, called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator), consists of a pool-type lead-cooled fast reactor, with a nominal thermal power of 300 MWt. This paper aims to verify the capability of the PHISICS/RELAP5-3D coupled approach to simulate transients of such reactor and to evaluate the effects of accidental scenarios relevant for the safety analysis on the system thermal-hydraulics and on the core power spatial evolution. RELAP5-3D©, developed at Idaho National Laboratory (INL), is a thermal-hydraulic system code, validated for a wide range of transient simulations. The code provides the possibility to simulate innovative working fluids (such as lead and lead alloys) and to use a fully integrated multi-dimensional nodalization. In addition, the need to study complex neutronic systems recommended the development of a new kinetic model allowing the calculation with any number of energy groups and also considering the transport for the angular variations. At this purpose, INL developed PHISICS (Parallel and Highly Innovative Simulation for INL Code System) and its coupling methodology with RELAP5-3D. The simulation activity described in this paper has been focused on the safety analysis of ALFRED reactor assuming the occurrence of two unprotected transient scenarios: unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP). At this purpose, a thermal-hydraulic geometrical scheme of the whole reactor has been developed. The models and the outcomes of the calculations are described and discussed in the paper, highlighting the capability of the coupling approach to obtain results consistent with the ones available in the literature.

The Water-Cooled Lithium-Lead Breeding Blanket is a key component of a fusion power plant, in charge of ensure Tritium production, shield Vacuum Vessel and magnets and remove the heat power deposited by radiation and particles arising from plasma. The last function is fulfilled by First Wall and Breeding Zone independent cooling systems. Several layouts of BZ coolant system have been investigated in the last years to identify a configuration that might guarantee EUROFER temperature below the limit (550 °C) and good thermal-hydraulic performances (i.e. water outlet temperature of 328 °C). A research activity is conducted to study and compare different modelling approaches to simulate the heat transfer within the BZ liquid metal, assessing their impact on the numerical prediction of the WCLL blanket thermal performances. An approach will rely on the simulation of convective and diffusive heat transfer processes taking place within the liquid metal by means of a CFD tool based on the Finite Volume Method. Conversely, the other approach will roughly assume a pure diffusive heat transfer mechanism within the breeder, due to the very low velocities envisaged for its flow field. In this case the heat transfer performances will be preferably assessed by means of a commercial code based on the Finite Element Method. The analyses have been carried out with reference to the so called “WCLL BB 2018 V0.6″ equatorial cell. Advantages and issues from the thermal-hydraulic point of view are identified, the impact of the imposed boundary conditions and heat transfer properties, with the implemented correlations, on the respective results is critically discussed.

The Advanced Lead Fast Reactor European Demonstrator (ALFRED) was conceived in the framework of the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project since 2010. Recently, a revised concept of the Reactor Coolant System (RCS) is ongoing to solve some identified issues such as thermal stratification and others. To verify the improvements of the ALFRED design the authors developed a detailed nodalization scheme of the reactor using RELAP5-3D© code. This paper provides a summary of the main improvements introduced in the revised concept of ALFRED reactor and a detailed discussion of the thermal-hydraulic analysis performed with RELAP5-3D©. The nodalization scheme is described in detail together with transient boundary conditions and timing of events. The numerical activity demonstrates the improvements of the revised configuration of the reactor that avoid the establishment of relevant thermal stratification phenomena in both normal and accidental operations. The effect of two relevant parameters is also investigated, highlighting that safety conditions are maintained in the whole spectrum of the sensitivity analysis.

In the Water-Cooled Lithium Lead (WCLL) blanket, the eutectic alloy lithium-lead (PbLi) is used as tritium breeder and carrier, neutron multiplier, and heat transfer medium. The PbLi hydraulic loop section in the range of the reactor field coils, which includes the blanket and a non-negligible length of the connection pipes, is affected by intense magnetic fields which cause the transition to a MHD regime. Lorentz forces oppose the fluid motion and cause pressure losses several orders of magnitude higher than for the ordinary hydrodynamic regime. An accurate estimate of the MHD pressure drop is mandatory to properly design the PbLi loop and to optimize the flow path in the blanket. In this paper, the so-called “in-magnet” section of the PbLi loop is divided into three main regions (feeding and draining pipe, manifold, and breeding zone) which are further discretized into basic hydraulic elements. Analytical correlations and numerical results available in the literature are then used to calculate the MHD pressure drop terms for each element and then the overall loss. The study highlights that the highest contribution to pressure loss is made by the spinal manifold and connection pipes. Optimization strategies are suggested to minimize losses in these elements.

In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering selected parameters as a figure of merit to predict possible code outcomes. The uncertainty band will be evaluated through sensitivity analyses programmed, collected and statistically manipulated through RAVEN software tool. MELCOR and RAVEN have been internally coupled through a new Python code interface developed by Sapienza University of Rome, to perform sensitivity and uncertainty quantification analyses during severe accident transient. The Beyond Design Basis Accident (BDBA) analysis of an ex-vessel loss of coolant accident (LOCA) for the water-cooled lithium lead (WCLL) blanket concept has been simulated with the fusion version of MELCOR 1.8.6 code. The postulated initiating event (PIE) is a double-ended break in the first wall (FW) cooling system distributor ring, with simultaneous failure of the plasma shutdown system. An in-vessel breach of the coolant system occurs because of FW failure, with consequent unmitigated plasma shutdown transient. Sensitivity analysis results have shown that the FW temperature at which plasma in-vessel breach occurs is strongly correlated with the mass of hydrogen produced. The same parameter has also an impact on the overall accident scenario, such as the trigger of VVPSS rupture disks and thus source term mobilization.

Abctract The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium production, shield the Vacuum Vessel and remove the heat generated by plasma thermal radiation and nuclear reactions. It relies on PbLi eutectic alloy adopted as breeder and neutron multiplier and refrigerate by subcooled pressurized water. The last function is fulfilled by two independent cooling systems: First Wall (FW) that faces the plasma heat flux and the Breeding Zone (BZ) that removes the deposited power of neutron and photon interaction. Several layouts of WCLL BB system have been investigated in the last years to identify a configuration that might guarantee EUROFER temperature below the limit (550 °C) and suitable thermal-hydraulic performances. This research activity focuses on the equatorial WCLL elementary cell of the Central Outboard Segment (COB), based on the WCLL BB 2018 design of DEMO 2017 baseline, investigating the cooling performances of the BZ and FW systems and their mutual interaction, verifying the reliability to deliver the coolant at adequate design conditions (i.e. 328 °C), evaluating the Eurofer temperature field in order to optimize the BZ Double Wall Tubes (DWTs), extinguishing the hot spot onset into the structures due to the implementation of the IAEA PbLi thermal conductivity as reference. The goal of the study is to compare three specific different tubes arrangement of the BZ system: the V0.6, that is the reference configuration, the V0.6_A and V0.6_B, identifying and discussing advantages and key issues from the thermal-hydraulics point of view. The analyses have been carried out using a CFD approach, adopting the commercial ANSYS CFX code, thus a 3D finite volume model of each configuration has been developed. For each WCLL design, several steady-state analyses have been performed, to evaluate the Eurofer temperature field and the thermal-hydraulics performances under different cooling configuration, changing the position of the DWTs. The results show that the FW and BZ system have a mutual interaction. Only the V0.6_B layout can safely remove the high heat loads deposited into the elementary cell, respecting the DEMO requirements. This research activity aims at laying the basis for the finalization of the WCLL BB design.

This work studies numerically the flow around an electrically insulated heating cylinder, bounded by walls of non-uniform electrical conductivity and subjected to a transversal magnetic field, with non-null components in the toroidal and poloidal directions. The configuration is representative of a typical breeding blanket segment in tokamak fusion reactors: to minimize magnetohydrodynamic (MHD) pressure drops, the liquid metal can be employed just as tritium breeder, whereas a non-conductive secondary fluid is used as coolant. The coolant is carried in the breeding zone by pipes that, being transversal to the stream-wise direction, affect the flow features and heat transfer. The flow is investigated by simulations performed in a 3D domain for Reynolds number 20 and 40, 0 ≤ M ≤ 50 for the Hartmann number and 0° ≤ α ≤ 32° for the magnetic field inclination on the toroidal axis. The transition to the MHD regime causes the suppression of the cylinder wake and the disappearance of the steady vortex structures. Electromagnetic coupling balances the flow rates between the top and bottom sub-channels, individuated by the cylinder. The flow pattern modifications affect the heat transfer, which is found to increase with both M and α in the considered range, albeit for the latter in a non-monotonic trend. The pressure drop in the channel exhibits a similar behaviour. Moreover, it is dominated by the fully developed component due to the 2D currents, whereas the cylinder pressure drop contribution decreases steadily with the intensity of the applied magnetic field. The simulations were performed with ANSYS CFX-15.

Chilled Water System (CHWS-1) is an ITER safety cooling system comprised of two independent and segregated trains (CHWS-1A and CHWS-1B), which provide cooling water to components located in Tokamak and Hot Cell Complexes. VV-PHTS is the only component that, during water baking mode, operates at high pressure than CHWS-1. In case of leaks, it represents a risk in terms of over pressurization and contamination of the CHWS-1 (due to tritiated water and activated corrosion products (ACPs)). Therefore, this paper postulates CHWS-1 leak events at one interface: Vacuum Vessel Primary Heat Transfer System. (VV-PHTS). Three scenarios have been analysed: an incident event (heat exchanger pipe leakage) and two accident events (heat exchanger tube break and multiple tubes break). A simplified RELAP5–3D model of CHWS-1B was developed to perform thermal hydraulic transient analysis as well as to define and size system operation and control. A postulated accident management strategy is included while also defining CHWS-1 instrumentation and control (IC it is not related to the actual CHWS-1B design. During VV-PHTS baking mode, decay HX is not foreseen to be operative. In order to avoid affecting the redundancy of the system, this study supports this decision.

The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to be used as a key component in fusion power devices. It provides tritium breeding and nuclear power conversion and extraction. Before its integration in a DEMO plant, a testing phase is required to gain data on both thermal and neutronic performances. For this purpose, a Test Blanket System is currently under conceptual design and will be integrated into dedicated ITER equatorial ports. Early insight on the safety performance of the Test Blanket System remains an essential element for integrating into ITER, and accident analysis is one of its critical components. Preliminary safety studies have been performed in the framework of Safety And Environment (SAE) work package of the EUROfusion consortium program to support the design and the integration of the Water-Cooled Lithium Lead Test Blanket System into ITER. At the present conceptual design phase, these studies are tailored to give insight on the effects of some parameters, such as postulated initiating event timing, plasma termination system or valves intervention, and possible safety provision implementation. A MELCOR model of the Water-Cooled Lithium Lead Test Blanket System has been developed, and two accident scenarios have been studied: an ex-vessel LOCA inside the port cell and a loss of flow accident because of pump seizure. The impact of mentioned parameters on accident evolution and consequences is investigated in support of the safety logic definition.

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.

Two-phase pressure drops in a mini-tube, in transition flow and subcooled boiling, are analyzed in the present paper, with the support of an experimental data set provided by ENEA with their BO.E.MI.A. test section. The methodology can be applied to different fluids according to similarity criteria. Single phase, subcooled and saturated conditions have been analyzed. The Reynolds number in the experiments was mainly in the transition zone between laminar and turbulent conditions, therefore a third order interpolation curve of the friction factor has been employed. The methodology is based on the model from Delhaye. The model considers the fluid properties, the energy, mass and momentum conservation equations to predict the ONB and OSV points and a hyperbolic function has been adopted to calculate the non-equilibrium vapor quality in the subcooled boiling region. The best agreement with the ENEA experimental data has been obtained using in the methodology the Chisholm correlation, with 83.59% of the predicted values with an error lower than 30%. © 2017 Elsevier Ltd

Water-cooled lithium-lead breeding blanket is considered a candidate option for European DEMO nuclear fusion reactor. ENEA and the linked third parties have proposed and are developing a multi-module blanket segment concept based on DEMO 2015 specifications. The layout of the module is based on horizontal (i.e. radial-toroidal) water-cooling tubes in the breeding zone, and on lithium lead flowing in radial-poloidal direction. This design choice is driven by the rationale to have a modular design, where a basic geometry is repeated along the poloidal direction. The modules are connected with a back supporting structure, designed to withstand thermal and mechanical loads due to normal operation and selected postulated accidents. Water and lithium lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium lead system. Rationale and features of current status of water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanics, thermo-hydraulics and neutronics analyses. Open issues and areas of research and development needs are finally pointed out.

The interaction between the molten metal and the plasma-containing magnetic field in the breeding blanket causes the onset of a magnetohydrodynamic (MHD) flow. To properly design the blanket, it is important to quantify how and how much the flow features are modified compared with an ordinary hydrodynamic flow. This paper aims to characterize the evolution of the fluid inside one of the proposed concepts for DEMO, the Water-Cooled Lithium Lead (WCLL), focusing on the central cell of the equatorial outboard module. A preliminary validation was required to gauge the capability of ANSYS CFX to deal with MHD problems. The buoyant and pressure-driven fully developed laminar flows in a square duct were selected as benchmarks. Numerical results were compared with theoretical solutions and an excellent agreement was found. The channel analysis was realized on a simplified version of the latest available design geometry, developed by ENEA, for M ≤ 1000. The simulation highlighted the formation of high velocity jets close to the baffle and the onset of an asymmetrical potential distribution.

Summary The water-cooled lithium–lead breeding blanket is a candidate option for the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium–lead as breeder–multiplier, pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015 specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach is employed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by means of radial–toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium–lead flows in a radial–poloidal direction. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operation and selected postulated initiating events. Water and lithium–lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium–lead system. Rationale and features of the single-module-segment water-cooled lithium–lead breeding blanket design are discussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration with the primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areas of research, and development needs are finally pointed out. @EUROfusion Consortium*, 2017. *For more details see http://www.euro-fusionscipub.org/disclaimer-copyright

Before the final shutdown of the PHÉNIX fast reactor, the CEA carried out a final set of experimental tests to gather data and additional knowledge on relevant sodium fast reactors (SFR) operation and safety aspects. One of these experiments conducted was the dissymmetrical configuration test, which was selected as benchmark transient within the H2020 SESAME project. ENEA and Sapienza University of Rome are participating in the benchmark using the RELAP5-3D© code. The thermal hydraulic analysis focuses on adequate core cooling prediction in accidental scenario. With the goal of investigating asymmetric thermal hydraulic behavior inside of the reactor pool, two different nodalization approaches have been applied for the RELAP5-3D model, which adopt the same geometrical scheme for the primary flow path, with the exception of the hot and cold pools and the core bypass. The first scheme has been developed using vertical parallel pipes with cross junctions for the hot and cold pools and an equivalent pipe to reproduce the core bypass. The second model includes a multidimensional (MULTID) component, which simulates the pools and provides a detailed nodalization of the core bypass. This study aims at assessing whether the two modeling approaches are equally capable to predict the asymmetrical temperature evolution over the test, caused by the azimuthal asymmetry of the boundary conditions. Blind calculation results are presented and discussed. The paper will be a first step toward the RELAP5-3D code assessment against the experimental results collected as part of the PHÉNIX dissymmetric test.

International audience; Phénix is a French pool-type sodium-cooled prototype reactor; before the definitive shutdown, occurred in 2009, a final set of experimental tests are carried out in order to increase the knowledge on the operation and the safety aspect of the pool-type liquid metal-cooled reactors. One of the experiments was the Dissymmetric End-of-Life Test which was selected for the validation benchmark activity in the frame of SESAME project. The computer code validation plays a key role in the safety assessment of the innovative nuclear reactors and the Phénix dissymmetric test provides useful experimental data to verify the computer codes capability in the asymmetric thermal-hydraulic behaviour into a pool-type liquid metal-cooled reactor. This paper shows the comparison of the outcomes obtained with six different System Thermal-Hydraulic (STH) codes RELAP5-3D©, SPECTRA, ATHLET, SAS4A/SASSYS-1, ASTEC-Na and CATHARE. The nodalization scheme of the reactor was individually achieved by the participants; during the development of the thermal-hydraulic model, the pool nodalization methodology had a special attention in order to investigate the capability of the STH codes to reproduce the dissymmetric effects which occur in each loop and into pools, caused by the azimuthal asymmetry of the boundary conditions. The modelling methodology of the participants is discussed and the main results are compared in this paper to obtain useful guide lines for the future modelling of innovative liquid metal pool-type reactors. © 2019 Elsevier B.V.

In the frame of the MYRRHA ADS technological development, an intense research effort has been undertaken by European Commission, in particular for the Heavy Liquid Metal technology assessment. In this scenario, EURATOM HORIZON2020 funded the SESAME & MYRTE projects, coordinating a series of thermal hydraulics experiments and simulations for the safety assessment of liquid metal fast reactors. In particular, in the MYRTE project, a dedicated activity has been defined, consisting of a low pressure experimental campaign realized at the ENEA Brasimone Research Centre and concerning a double wall Steam Generator Bayonet Tube (SGBT), aiming at supporting the development of the Primary Heat Exchanger (PHX) of MYRRHA. The CIRCE (CIRColazione Eutettico) pool facility hosts a devoted Test Section (TS) named HERO (Heavy liquid mEtal pRessurized water cOoled tubes). HERO consists of seven double wall bayonet tubes, with an active length of 6 m, arranged in a hexagonal geometry. This solution aims at improving the plant safety reducing the possibility of water-lead/leadalloy interaction thanks to a double physical separation between them and allowing an easier control of potential leakages of the coolant by pressurizing the separation region with inert gas. The experimental campaign in CIRCE-HERO has been supported by a pre-test analysis using system thermal-hydraulic codes (i.e. RELAP5). A sensitivity analysis has been performed considering the mass flow rate and HERO inlet temperature for both the water and LBE, fixing the operating pressure at 16 bar. A test matrix, as produced as output of the sensitivity analysis and aimed at reproducing as faithfully as possible the thermal-hydraulic behaviour of the MYRRHA PHX under different operating conditions, is here presented. The experimental results collected from the low pressure tests on HERO test section are reported and discussed.

The conceptual design of the Primary Heat Transfer System (PHTS) of the water-cooled European (EU) DEMO foresees two independent cooling circuits, the breeding zone PHTS and the first wall PHTS. During the pulse time (120 min) the first delivers thermal power to the turbine, the latter delivers thermal power to the Intermediate Heat Transfer System (IHTS) equipped by an Energy Storage System (ESS). The IHTS delivers partially thermal power to the turbine in pulse so that to accumulate a suitable amount of energy in ESS to operate the turbine during the dwell time (10 min) at almost constant load, despite the EU DEMO pulsation of the generated thermal power. A dynamic model of the primary systems of water-cooled EU DEMO is developed using RELAP5/Mod3.3 system code to verify components sizing and to investigate code predictive capabilities. The model includes the primary and secondary side of the breeding zone and first wall, and in particular: in-vessel (i.e. breeding blanket) and ex-vessel components (i.e. main collectors, hot and cold legs, heat exchangers, steam generators, pumps). Preliminary assessments of the nodalization have been carried out, in particular checking pressure drops along the systems and heat exchanger performances.