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Transient analyses in a water-cooled fusion DEMO (Demonstration Power Plant) reactor have been performed to support the WCLL (Water-Cooled Lithium Lead) breeding blanket design. In this framework, the Design Basis Accident (DBA) analysis of an in-box LOCA has been carried out. The WCLL breeding blanket concept relies on Lithium Lead (LiPb) as breeder, neutron multiplier and tritium carrier, which is cooled by water at 15.5 MPa with an inlet temperature of 295 °C and an outlet temperature of 328 °C. Water flows in Double-Wall Tubes (DWTs) in order to reduce the probability of water/LiPb chemical interaction. In the case of a LOCA multiple rupture of these tubes is postulated, with consequent leakage of pressurized water in the LiPb side of the module. The analysis has been performed with the MELCOR computer code (ver. 1.8.6) modified for the application in the fusion context. Custom models are employed to simulate the chemical water/LiPb interaction in the module. The rupture mass flow rate calculated in water simulation is transformed in its equivalent in terms of hydrogen and unreacted water steam. Both have been treated as non-condensable gas. Two different input decks, one for each fluid considered, have been coupled through an external interface to account for their reciprocal interaction. Pressure and temperature transient behavior in the broken module demonstrate that safety margins are respected during the whole accidental sequence, even though no external safety system is foreseen or actuated. Moreover, attention has paid to the quantity of hydrogen produced in order to support the development of solutions, suitable for DEMO, to prevent hydrogen explosion.

The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium self-sufficiency, shielding the Vacuum Vessel and removing the generated power of DEMO reactor. The last function is fulfilled by two independent cooling systems: First Wall (FW) that removes the plasma heat flux and Breeding Zone (BZ) that removes the deposited power of charged particle and neutron. Several layouts of BZ coolant system have been investigated in order to identify a configuration that guarantee Eurofer temperature below the limit (823 K) and good thermal-hydraulic performances (i.e. water outlet temperature 601 K). A research activity is conducted to study and compare four configurations, which rely on different arrangement of the stiffening plates (i.e. toroidal-poloidal and radial-poloidal), orientation of the cooling pipes (i.e. horizontal, vertical) and PbLi flow path. The analysis is performed using a CFD codes, thus a 3D finite volume model of each configuration is developed, adopting the commercial ANSYS CFX code. The objective is to compare the BZ cooling system layouts, identifying and discussing advantages and key issues from the thermal-hydraulic point of view, also considering feedbacks from MHD and neutronic analyses. The research activity aims at laying the groundwork for the finalization of the WCLL blanket design, pointing out relevant thermal-hydraulic aspects.

The water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European 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. Current design is based on DEMO 2017 specifications. Two separate water systems are in charge of cooling the first wall and the breeding zone: thermo-dynamic cycle is 295–328 °C at 15.5 MPa. The breeder enters and exits from the breeding zone at 330 °C. Cornerstones of the design are the single module segment approach and the water manifold between the breeding blanket box and the back supporting structure. This plate with a thickness of 100 mm supports the breeding blanket and is attached to the vacuum vessel. It is in charge to withstand the loads due to normal operation and selected postulated initiating events. Rationale and progresses of the design are presented and substantiated by engineering evaluations and analyses. Water and lithium lead manifolds are designed and integrated with the two consistent primary heat transport systems, based on a reliable pressurized water reactor operating experience, and six lithium lead systems. Open issues, areas of research and development needs are finally pointed out. © 2019 Elsevier B.V.

The water-cooled EU-DEMO Breeding Blanket (BB) is cooled by two independent circuits, the Breeder Zone (BZ) and the First Wall (FW) Primary Heat Transfer Systems (PHTS). The configuration under study foresees the presence of an Intermediate Heat Transfer System and an Energy Storage System to operate the turbine during both the pulse (2 h) and the dwell time (10 min) at almost constant load, despite the plasma power pulsation. Within the framework of the EUROfusion WPBOP research activity, a RELAP5/Mod3.3 model was developed to investigate the thermal-hydraulic behavior of the primary cooling systems during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”. The nodalization includes the BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat EXchangers. The model was initially used to simulate the nominal conditions with the pulse and dwell phases. Then, starting from the pulse, a Loss of Flow Accident (LOFA) was selected to preliminary evaluate the PHTS behavior with the aim of the design improvement. LOFA analyses were performed considering the complete loss of both the FW and BZ PHTS main coolant pumps (MCPs). A sensitivity was carried out to assess the impact of the MCPs flywheel on the main PHTS parameters. Transient results highlighted the appropriateness of the current design with no need for further mitigation actions.

Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

In the frame of EUROfusion Work Package Breeding Blanket, thermal-hydraulic engineering analyses were carried out at ENEA Brasimone, in close cooperation with Sapienza University of Rome, to investigate thermal and fluid-dynamic behavior of the Water Cooled Lithium Lead Breeding Blanket, which is a candidate option for the European DEMO nuclear fusion reactor. The research activity focused on the equatorial elementary cell of an outboard segment with the aim of obtaining a temperature map of breeding blanket domains and to evaluate breeding zone and first wall systems cooling ability. A three-dimensional finite volume model of the breeding unit was developed, adopting the commercial CFD code ANSYS CFX 15.0. Steady state simulations were carried out considering a heat flux of 0.5 MW m−2, a radial power density distribution applied to all domains, and buoyancies forces activated in PbLi region. Results show that FW and BZ systems have good cooling efficiency, maintaining the maximum temperature in PbLi and solid structures well below 500 °C. The thermal-hydraulic results are discussed, highlighting open issues and suggesting pertinent modifications to DEMO WCLL coolant system layout aimed at optimizing the design.

The Water-Cooled Lithium Lead breeding blanket is a candidate option for the realization of European DEMO power plant. One of the main functions of the breeding blanket is to recover the thermal power from the first wall and the breeding zone and to drive it to the Primary Heat Transfer System. Moreover, due to the DEMO pulsed operation, an Energy Storage System is foreseen in order to ensure thermal energy availability and reduce cycling loading during dwell time. The blanket design must interface with the Primary Heat Transfer System and the Intermediate Heat Transfer System in an integrated solution with the aim of delivering feasible power plant, with high conversion efficiency. A thermal-hydraulic model of the main components of the water cooled Primary Heat Transfer System is developed, using RELAP5 system code, to support the design and sizing. A RELAP5/Mod3.3 extended version has been set-up to implement PbLi and molten salt (HITEC) fluid proprieties, as well as the relevant heat transfer correlations. The results of the preliminary applications provide a comprehensive set of thermo-hydraulic parameters (i.e. coolant inventory, pressure drops), relevant for the assessment of the design integration.

In the present paper, the possibility of increasing the thermodynamic efficiency of an electric energy production plant, by using an advanced energy conversion system based on supercritical carbon dioxide (S-CO2) as working fluid, has been analyzed. Since the supercritical carbon dioxide cycles are being considered as a favorable candidate for the next generation of nuclear power plant energy conversion systems, a lead cooled fast reactor has been selected as reference in the present analyses. The main aim of the present study is to compare two different S-CO2 thermal cycles applied on the conversion system of a nuclear power plant. The reference Lead cooled Fast Reactor (LFR) used for the present analyses is the ALFRED reactor, which has a thermal power of 300 MW and it is considered the scaled down prototype of the industrial European Lead Fast Reactor (ELFR). Thermodynamic cycles selected for the present study are a Recompression Cycle and a Brayton Cycle with Regeneration. Each of them has been analyzed under several design conditions regarding the maximum pressure and the regeneration coefficient. Among different design conditions, the solution allowing the maximization of the overall efficiency has been identified. Thermodynamic analyses have been carried out with GateCycle™ v. 6.1.1, which is a General Electric software able to predict design and off-design performance of power plants.

The U.S. Nuclear Regulatory Commission is currently proposing rulemaking to revise the Loss Of Coolant Accident (LOCA) and therefore the Emergency Core Cooling System (ECCS) acceptance criteria, to include the effects of higher burnup on cladding performance as well as to address other technical issues. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost, as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The study here presented has been part of a big project used to investigate technical issues and approaches for future industrial applications within the Risk-Informed Safety Margin Characterization (RISMC) Pathway. Specifically, the primary aim of this study is to lay out a roadmap to demonstrate the application of the new methodology. The present analysis shows a simplified version of the methodology of an industrial application on the Core Design and the Multi-Cycle Analysis.

Magnetohydrodynamic pressure drops are one of the main issues for liquid metal blanket in fusion reactors. Minimize the fluid velocity at few millimeters per second is one strategy that can be employed to address the problem. For such low velocities, buoyant forces can effectively contribute to drive the flow and therefore must be considered in the blanket design. In order to do so, a CFD code able to represent magneto-convective phenomena is required. This work aims to gauge the capability of ANSYS© CFX-15 to solve such cases. The laminar flow in a differentially heated duct was selected as validation benchmark. A horizontal and uniform magnetic field was imposed over a square duct with a linear and constant temperature gradient perpendicular to the field. The fully developed flow was analyzed for Gr = 105 and Hartmann number (M) ranging from 102 to 103. Both insulating and conducting duct walls were considered. Strong dampening of the flow in the center of the duct was observed, whereas high velocity jets appeared close to the walls parallel to the magnetic field. The numerical results were validated against theoretical and numerical results founding an excellent agreement.

[EN] The MOSCAB experiment (Materia OSCura A Bolle) uses the "geyser technique", a variant of the superheated liquid technique of extreme simplicity. Operating principles of the new dark matter detector and technical solutions of the device are reported in detail. First results obtained in a series of test runs taken in laboratory demonstrate that we have successfully built and tested a geyser-concept bubble chamber that can be used in particle physics, especially in dark matter searches, and that we are ready to move underground for extensive data taking., We gratefully acknowledge support from Istituto Nazionale di Fisica Nucleare (INFN) and Spanish Ministerio de Economia y Competitividad (MINECO), Grants FPA-2015-65150-C3-2-P (MINE-CO/FEDER), and Consolider MultiDark CSD2009-00064. We are grateful to Laboratori Nazionali delGran Sasso (LNGS) for hosting and supporting the MOSCAB project. We thank Massimo Locatelli, Tecnologia Meccanica, Treviolo (BG), Italy, for supporting our project with hard work, competence and passion, and Francesco Boccardo, Vincenzo Caglioti and Dario Menasce for fruitful discussions and their contribution in preparing the manuscript.

Considering that solutions for the steady-state power exhaust problem in future fusion reactors (e.g. DEMO) are not provided by present experiments and it is uncertain if they will be provided by ITER, because the expected heat fluxes, as well as the level of neutron irradiation, will be much higher, dedicated work packages are being devoted to this problem within EUROfusion and even a dedicated facility (the Divertor Tokamak Test − DTT) is being proposed in Italy. Among the possible options, a liquid metal (LM) divertor is being considered. The present work aims at developing a simple model of the LM loop in the case of a pool-type divertor, including the most important physical phenomena and allowing to roughly determine the operating range of the system, in terms of surface temperatures and vapor pressures. This work therefore sets a preliminary basis for the conceptual design of a LM divertor for the DTT facility. The model includes the incoming plasma heat load and a basic treatment of the interactions of the Li vapor with the plasma. The reduction of the Li vapor efflux due to ionization by the plasma is also taken into account. The model includes two chambers: a first divertor box, the evaporation chamber (EC), is open towards a second divertor box, the differential chamber (DC), which is in turn connected to the main plasma chamber (MC). The model is used to study the effectiveness of the LM vapor in radiating isotropically the parallel heat flux incoming in the divertor. The results indicate that the presence of the DC allows a significant reduction of the Li vapor efflux towards the MC, which in turn would imply a lower contamination of the core plasma.

In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R&D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

The objective of this experimental study is to evaluate the heat transfer coefficient outside a tube with annular transverse fins, derived from strips of copper mechanically bound and coupled outside. Water is used as the heating medium, in turbulent conditions and flowing at different temperatures inside the tube. Petukhov's correlation has been selected to calculate the water heat transfer coefficient in the tube. The experimental data obtained are compared with a correlation from literature, and a similar trend is observed. A fitting of the data provides a correlation for the three tubes of different external diameter (30 mm, 22 mm, and 15.6 mm) that agrees very well with the experimental values. The thermal contact conductance is identified as the main reason for the difference between data and the original Briggs and Young correlation. An estimation of the contact conductance between fins and tubes provides values between 3500 and 11000 W/m2-K, slightly increasing with the air Reynolds number (based ...

A theoretical and experimental investigation on steam condensation in presence of noncondensable gases within horizontal and inclined tubes is summarized in the present paper. A simple correlation mainly based on dimensionless numbers was derived and compared with previous formulations based on the diffusion layer model. The noncondensable gases presence during condensation is an important issue affecting the whole thermodynamic efficiency of the process, and for this reason highly investigated by many researchers. The experimental data obtained for condensation, inside horizontal or inclined tube (15°, 30° and 45°) with an internal diameter of 22 mm, of an air/steam mixture, at low mixture Reynolds numbers (

Free surface vortices formation has been, for many years, a relevant issue in many engineering applications with hydraulic intakes, since strong whirlpools introduce swirl flow at the intake causing entrainment of floating matters and gases (air or other gases). The Gas Entrainment (GE) phenomenon has also been an important topic in nuclear industry in the last years, due to the possibility of free vortex formation at the surface, especially in sodium-cooled fast reactors (SFRs). This phenomenon may result in unlikely positive reactivity insertion accident, affecting the safety performances of the reactor itself. The GETS experimental facility (Gas Entrainment Test Section) has been built in the thermal–hydraulics laboratory of the DIAEE in order to study free surface vortices occurrence. The main purpose of this facility is to understand the influence of different parameters on free surface vortices formation and evolution. Experimental tests and preliminary observations, carried out, with cold water as working fluid, different water level in the tank, circulating flow rate and outlet tube diameter, are presented as occurrence maps in this paper. The observed vortices have been qualitatively classified in different “formation stages” and the most important parameters that influence the physical phenomenon have been identified through a dimensionless analysis. Empirical correlations based on dimensionless parameters of the transition boundaries between stages, including the onset of gas entrainment, have been developed and discussed, and very preliminary considerations on the sodium–water similitude, based on the Froude and Weber numbers, are introduced.

A numerical activity, aimed to evaluate the capability of RELAP5-3D© to reproduce the main thermal-hydraulic phenomena in an HLM pool-type facility, in different operative conditions, is presented. For this purpose, the experimental campaign performed in CIRCE-ICE test facility has been selected for the code assessment. Two experimental tests have been analyzed: TEST A consisting in a transition from no-power to a full power steady state conditions, and TEST I, consisting in a transition from gas-enhanced circulation to natural circulation, simulating a protected loss of heat sink plus a loss of flow accident. Three different pool modelling approaches are presented, consisting in a single vertical pipe, parallel pipes with cross junctions and multi-dimensional component. The comparison with experimental data has highlighted the need to divide the large pool in several sections to reproduce the natural convection, strictly correlated with the thermal stratification. The multi-dimensional component seems to be the best practice for the evaluation of this phenomenon even if the lack of specific correlation for heat transfer coefficient in quasi-stagnant conditions in large tanks is a limit for the accuracy of the results. In addition, the paper presents a detailed nodalization of the fuel pin bundle, highlighting quite good capabilities of RELAP5-3D as a subchannel analysis code.

An experimental investigation on the role of noncondensable gases during condensation of steam inside inclined tubes is presented. In a condenser, noncondensable gases flowing with steam cause reduction of condenser performance and efficiency. Many researchers have investigated in-tube condensation for vertical heat exchangers, but very few works have been performed to study condensation in inclined tubes with noncondensable gases. In the paper, experiments dedicated to this situation are described, with reference to the following conditions: tube internal diameter: 12.6 mm, 20 mm and 26.8 mm; tube inclination: 7°, 15°, 30° and 45°; inlet noncondensable gas mass fraction ωin = 2%–42%; inlet mixture Reynolds number Rem,in = 5000–20000; local noncondensable gas mass fraction ω = 2%–70%; local mixture Reynolds number Rem = 400–21000; local condensate Reynolds number Rel = 10–290; saturated steam at atmospheric pressure; and gravity controlled flow regime. A limited influence of the inclination angle on heat transfer coefficient has been observed. Correlations to evaluate the local heat transfer coefficient along inclined tubes, in a gravity controlled flow regime, have been developed and they are in good agreement with the experimental results.

An experimental investigation on the role of noncondensable gases during condensation of steam inside horizontal/slightly inclined tubes is presented. In a condenser of a thermal desalination unit, the noncondensable gases flowing within the gas phase cause reduction of performance and efficiency. Many researchers have investigated in-tube condensation for horizontal heat exchangers, but very few works have been performed to study the condensation in slightly inclined tubes with noncondensable gases. Inclination of tubes may be requested to achieve suitable and reliable draining. Experiments were conducted in the following conditions: tube internal diameter 12.6 mm, 20 mm and 26.8 mm; tube inclination 7°; inlet noncondensable gas mass fraction ωin = 5%–42%; inlet mixture Reynolds number Rem,in = 5000–20,000; local noncondensable gas mass fraction ωin = 5%–60%; local mixture Reynolds number Rem = 500–20,000; saturated steam at atmospheric pressure, gravity controlled with stratified flow regime, in which the condensate is collected mainly in the bottom part of the tube due to gravity and it is drawn out by its own momentum. A correlation of mixed gas heat transfer coefficient along a slightly inclined tube, in a gravity controlled flow regime, has been developed, showing a good agreement with experimental results.

An experimental investigation on bathtub vortices has been performed by using Particle Image Velocimetry (PIV). Velocity fields associated to the free surface vortex were obtained at three horizontal planes and four Reynolds numbers, i.e. between 2400 and 11,000 (calculated with reference to the exit hole diameter and the mean exit velocity). Due to the unsteady behavior of the flow field, the vortex center positions have been identified and the vortex paths were reconstructed for all experiments. Average velocity fields have been calculated by aligning the vortex centers at each frame, in order to derive radial and tangential velocity profiles, to be compared at different Reynolds numbers and measurement planes. The results show that the radial motion assumes a potential behavior when it is near the exit hole, scaling quite well with the average exit velocity (thus with the corresponding Reynolds number). On the other hand, the tangential component is well approximated by the Rankine’s flow potential solution only near the free surface, the tangential velocity peak increment not being linearly proportional to the outlet velocity. Vorticity fields and circulation profiles have been derived from the measured velocity fields and discussed. Turbulence fluctuations statistical analysis gives also evidence of a clear dependence on Reynolds number and distance from the exit hole.

The latest versions of RELAP5-3D©code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D©water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D©default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential), specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead), were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

Due to the important role of microscale heat transfer, an analysis of the heat transfer coefficient for a 1 mm inner diameter tube, with FC-72 as working fluid, has been performed. The study is aimed to identify the best correlation or model to predict the available experimental database. A comparison of several models and correlations, available in the literature for micro- and macroscale, has been performed, focusing the present preliminary analysis to the saturated boiling conditions. Experimental data have been provided by ENEA through the facility BO.E.MI.A. (BOiling Experiments in MIcrochannel Apparatus), in the pressure range from 3 to 5 bar, with a mass flux from 800 to 1200 kg/m 2 s and thermal fluxes from 1.6 to 181 KW/m 2 . The best results, in this preliminary analysis of the saturated boiling points, were obtained for microscale correlations of Li and Wu, with more than 91% of data within ± 30% error and a mean absolute percent error (MAPE) of 13.4%. Among the macroscale correlations only the Chen correlation provides good results, but with a lower agreement with the experimental data. Theoretical models are very promising but need further works to find the appropriate parameters valid for the specific fluid. A more detailed analysis including subcooling conditions will be performed in a future work.

In a Tokamak-fusion reactor, the plasma interacts with the first wall generating dust of different materials. Dust can play an important, if not primary, role in determining safety and economic performance of this type of fusion machines, due to potential accident and maintenance requirements. In this paper, after a brief introduction, a description of dust explosion mechanism is reported with reference to several experiments used to create or validate theories and models. The safety concerns about dust are based on the assumption that, in case of air inflow events, this particulate can be resuspended in the vacuum vessel leading to a potential explosive cloud. In the paper the main literature models used to deal with dust transport and resuspension are reported: the VZFG model, the Rock n’ Roll model and the ECART model. Differences, pros and cons of each models are described to obtain a comparison between energy-based and force-based balance models. Finally, the numerical approach for the resuspension phenomenon is discussed and an application is shown. The simulation has been carried out with reference to a laboratory scale facility-like in 3D geometry. The simulated event is an air flow intake, as from a Loss of Vacuum Accident, and the results show the distribution of dust in the geometry after the first instants of intake, the mass concentration of particulate on the surfaces and the path lines of the flow field. The simulated geometry is only partially referred to an existing experimental facility, and the methodology could be useful to replicate the same conditions to obtain a validation of the results. More and finer simulations are foreseen to reproduce more accurately experimental data and real scenarios, with the aim to evaluate the risk of explosion by means of an accurate prediction of dust distribution inside a vacuum volume during resuspension due to air inflow.

The design of fusion devices often includes water as primary coolant of the first wall/blanket system and a potential accident scenario is the steam/water injection from the primary circuit into the magnets' cryogenic chamber (cryostat). An important question to be answered for the above scenario is whether the pressure suppression created by the cryogenic surfaces is sufficient to prevent cryostat damage. The computer codes used for the assessment of ITER safety were validated in the past years against the EVITA (Experimental Vacuum Ingress Test Apparatus) experiment at CEA-Cadarache, which was designed for the simulation of the physical phenomena occurring during a coolant ingress into the cryostat, namely ice formation on a cryogenic structure, heat transfer coefficient between walls and fluid, flashing, two-phase critical flow. The paper presents the results obtained by the CONSEN computer program for seven post-test calculations of the EVITA facility relating to the cryogenic experiments carried out in 2004 and 2005, in which the kinetics of the ice layer formation was analysed. The comparison with the experimental data has been performed and the main agreements and differences are remarked. © 2014 Elsevier Ltd.

The Supercritical carbon dioxide (SCO2) is widely used in many industrial applications as process or heat transfer fluid. The SCO2 is characterized by good chemical stability at high temperature, supercritical conditions at limited temperature and pressure and high thermodynamic efficiency. Because of these features, SCO2 allows to operate with small sized operating machines and higher efficiency compared to conventional applications. Moreover, the SCO2 can be used in conventional thermodynamic gas cycles reaching higher efficiency at the same temperature level. The main advantage is the possibility of reaching thermodynamic efficiency values close to 50%, much higher than conventional cycles. Moreover, the critical point of CO2, low pressure and low temperature (7 MPa and 300 K), makes systems more economic and technically feasible than systems powered by other supercritical fluids. SCO2 industrial applications are analyzed in the present paper and specific advantages will be investigated.

A CFD analysis has been carried out to study the thermal-hydraulic behavior of liquid metal coolant in a fuel assembly of triangular lattice. In order to obtain fast and accurate results, the isotropic two-equation RANS approach is often used in nuclear engineering applications. A different approach is provided by Non-Linear Eddy Viscosity Models (NLEVM), which try to take into account anisotropic effects by a nonlinear formulation of the Reynolds stress tensor. This approach is very promising, as it results in a very good numerical behavior and in a potentially better fluid flow description than classical isotropic models. An Anisotropic Shear Stress Transport (ASST) model, implemented into a commercial software, has been applied in previous studies, showing very trustful results for a large variety of flows and applications. In the paper, the ASST model has been used to perform an analysis of the fluid flow inside the fuel assembly of the ALFRED lead cooled fast reactor. Then, a comparison between the results of wall-resolved conjugated heat transfer computations and the results of a decoupled analysis using a suitable thermal wall-function previously implemented into the solver has been performed and presented.

The Loss of Coolant Accident in a Vacuum Chamber is one of the safety issues for the fusion facilities. With the aim of validating computer codes able to simulate thermal-hydraulic transient events in fusion reactors, experimental tests were carried out for comparing experimental data with simulations results. Among others, the Integrated ICE test facility (JAERI) has been developed to reproduce the ITER reactor geometry with a scaling factor of 1/1600. In order to perform safety assessment of fusion reactors, several computer codes have been developed and among them CONSEN, a fast running computer code developed to simulate thermal-hydraulic transients. The present paper is aimed at identifying the CONSEN capability to reproduce the behavior of main physical parameters during an Ingress of Coolant Event and consequent loss of vacuum in the Plasma Chamber of a fusion reactor. In particular, the CONSEN simulations results are here compared with experimental data obtained at the modified Integrated ICE test facility during the tests P1 and P7. Simulation results and experimental data have been compared and a good agreement has been identified for most of analyzed parameters. © 2015 Elsevier B.V. All rights reserved.

The OECD/NEA proposes a new international benchmark based on data collected during an instability transient occurred at the Oskarshamn-2 NPP. This benchmark is aimed at testing the coupled 3D Neutron Kinetic-Thermal Hydraulic (3D NK-TH) codes on challenging situations. The ENEA "Casaccia" Research Center, is participating to this benchmark, developing a computational model using the RELAP5-3D© code. The 3DNK model has been already developed from the cross sections dataset calculated by OKG, the Oskarshamn-2 licensee, through the CASMO lattice code. In order to use this neutron cross sections database in RELAP5-3D© a n-dimensional polynomials data fitting and base cross sections values calculations are required. An ad-hoc tool, named PROMETHEUS, has been developed for automatically generating RELAP5-3D©-compatible cross sections libraries. This tool allows at easily visualizing the complex structure of the neutronic datasets; moreover it is exploited for deriving different cross sections libraries needed to evaluate neutronic parameters effects on the reactor instability prediction. Thus, the effects of the fuel temperature and control rod histories, of the discontinuity factors (averaged/not averaged) and of the neutron poisons has been assessed. A ranking table has been produced, demonstrating the relevance of the not-averaged discontinuity factors and of the on-transient neutron poisons calculations for the correct prediction of the Oskarshamn-2 event. © 2015 Elsevier Ltd. All rights reserved.

In the frame of heavy liquid metal (HLM) technology development, CIRCE pool facility at ENEA/Brasimone Research Center was updated by installing ICE (Integral Circulation Experiments) test section which simulates the thermal behavior of a primary system in a HLM cooled pool reactor. The experimental campaign led to the characterization of mixed convection and thermal stratification in a HLM pool in safety relevant conditions and to the distribution of experimental data for the validation of CFD and system codes. For this purpose, several thermocouples were installed into the pool using 4 vertical supports in different circumferential position for a total of 119 thermocouples [1][2].The aim of this work is to investigate the capability of the system code RELAP5-3D© to simulate mixed convection and thermal stratification phenomena in a HLM pool in steady state conditions by comparing code results with experimental data. The pool has been simulated by a 3D component divided into 1728 volumes, 119 of which are centered in the exact position of the thermocouples. Three dimensional model of the pool is completed with a mono-dimensional nodalization of the primary main flow path. The results obtained by code simulations are compared with a steady state condition carried out in the experimental campaign. Results of axial, radial and azimuthal temperature profile into the pool are in agreement with the available experimental data Furthermore the code is able to well simulate operating conditions into the main flow path of the test section.

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Hydrogen leakage and fire ignition and propagation are safety concerns in several industrial plants. In a nuclear fusion power plants the separation of hydrogen and tritium takes place in different steps, among which one or more electrolyzers are foreseen. A fire scenario could take place in case of leakage of hydrogen. In such cases, it is important to prevent the spreading of the fire to adjacent rooms and, at the same time, to withstand the pressure load on walls, to avoid radioactivity release in the surrounding environment. A preliminary study has been carried out with the aim of comparing CFD tools for fire scenario simulations involving hydrogen release. Results have been obtained comparing two codes: ANSYS Fluent© and FDS. The two codes have been compared both for hydrogen dispersion and hydrogen fire in a confined environment. The first scenario is aimed to obtaining of volume fraction 3D maps for the evaluation of the different diffusion/transport models. In the second scenario, characterized by a double-ended guillotine break, the fire is supposed to be ignited at the same time of the impact. Simulations have been carried out for the first 60 seconds. Hydrogen concentration, temperature and pressure fields are compared and discussed.

For the future development of Generation IV nuclear reactors, both safety and economic targets have to be achieved. In order to increase, at the same time, the power density generation and the safety features, a huge R&D effort is still required. Referring especially to Liquid Metal Cooled Fast Reactors, much attention is placed on Gas Entrainment (GE) phenomena, which could cause unlikely positive reactivity insertion accident. The GETS experimental facility (Gas Entrainment Test Section), especially aimed at studying the free surface vortices occurrence, has been built in the thermal-hydraulics laboratory of the DIAEE. The main purpose of this facility is to identify the most important parameters affecting the whirlpools formation and evolution. Experimental tests and preliminary observations have been performed. Different vortex behaviours related to different experimental conditions have been identified and presented in the present paper. 2D occurrence maps as function of different dimensionless groups (Reynolds, Froude and Weber numbers and H* = H/d ratio) have been defined. In the present paper, the results of a first experimental campaign, carried out with tap water, are discussed.

The project concerns the design, optimization, construction, assembling, start-up and extensive monitoring of a full-titanium desalinator. The operational tests took place at the site of the solar pond of the University of Ancona (Italy). Data collected during manufacturing tests and, in particular, during the start-up and the operation of the plant under various conditions, are being utilized for improving expertise on heat recovery with highly corrosive fluids, on co-generation plants aimed at producing electricity and fresh water, and on desalination fed by solar energy. One major output of the project is the assessment of data which permits an evaluation of the additional cost of the full-titanium desalinator with respect to a traditional technology, with the added benefits of (a) better heat transmission through tube bundles, which means better performance; (b) a reduced need for chemicals and maintenance activities; and (c) improved plant reliability and duration. The research program has three main aims: (1) the improvement of the economics of solar desalination, namely desalination of water through operation of solar ponds; (2) the demonstration of thermal performance, maintenance and chemical requirements, reliability and overall costs of a full-titanium desalination plant through operation of a plant of meaningful size in order to disseminate the technology of full-titanium desalination plants in the electric-energy production industry for use in co-generation units; and (3) the improvement of knowledge regarding industrial-size use of heat recovery from highly corrosive fluids.

The paper analyses the pressurization of the ITER cryostat during an ex-vessel loss of coolant accident (LOCA) in the first wall–shielding blanket (FW–SB) primary cooling loop. Since the cryostat is a strong barrier against the release of mobilized hazardous material, its integrity is essential in meeting the ITER safety objectives. The analyses refer to ITER TAC 4 design and were performed by means of a fast running thermal fluid dynamic code that allowed parametric studies. Helium LOCA can be handled by the code, so that the analyses did take into account phenomena such as coil quench and electrical arcing between coil turns, that may occur in tokamaks with superconducting cables in the magnet coils. The main result was that even in the worst conditions the pressure transient in the cryostat does not damage its integrity provided that the plasma facing components (PFC) cooling loops in the vacuum vessel remain undamaged. Hence, if no other energy sources than those coming from an ex-vessel LOCA are added to water, helium and heat structures in the cryostat, the cryostat maintains its integrity. More analyses are needed to evaluate an in-vessel LOCA and the consequent cryostat pressurization transient. In fact this event would add decay heat and exothermic chemical reaction as energy sources to be possibly taken into account in the pressurization transient. These analyses are not within the scope of this work.

The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code. © 2013 Published by Elsevier B.V.

Code validation activities have been promoted inside the European fusion development agreement (EFDA) to test the capability of codes in simulating accident phenomena in fusion facilities and, specifically, in the International thermonuclear experimental reactor (ITER). This work includes a comparison between three different computer codes (CONSEN, MAGS and MELCOR) and one analytical model (ITER Model) in simulating cryogenic helium releases into the vacuum vessel (VV) which contains hot structures. The scope was the evaluation of the transient pressure inside the VV. The results will be used to design a vent duct (equivalent diameter, length and roughness) to allow pressure relief for the protection of the VV, which has a maximum design pressure of 200 kPa. The model geometry is a simplified scheme preserving the main features of the ITER design. Based on the results of the simulations, a matrix of experiments was developed to validate the calculated results and to design the vent duct for the ITER VV. The experiments are planned to be performed in the EVITA test facility, located in the CEA Cadarache research centre (France).

In the frame of the deterministic approach in the safety analysis of the ITER reactor, it is postulated that a helium cooling line from the cryogenic system and one coolant pipe of the first wall/blanket primary heat transport system are damaged simultaneously in the cryostat. This event is selected to bound possible damage to the systems in the cryostat. The computer codes, which are used for the assessment of ITER safety, can be validated against the EVITA experiment, which has been designed for the simulation of the physical phenomena occurring during a coolant ingress into the cryostat, namely ice formation on a cryogenic structure, heat transfer coefficient between walls and fluid, flashing, two-phase critical flow. The comparison between calculations and experiments allows the ability of the computer codes to treat the relevant physical phenomena to be assessed.

This paper summarises our recent investigations undertaken as part of the EURODESAL project on nuclear desalination, currently being carried out by a consortium of four European, and one Canadian, industrials and two leading EU R&D organisations. Major achievements of the project, as discussed in this paper are: • Coherent demonstration of the technical feasibility of nuclear desalination through the elaboration of coupling schemes for optimum cogeneration of electricity and water and by exploring the unique capabilities of the innovative nuclear reactors and desalination technologies. • Verification that the integrated system design does not adversely affect nuclear reactor safety. • Development of codes and methods for an objective economic assessment of the competitiveness and sustainability of proposed options through comparison, in European conditions, with fossil energy based systems. Results obtained so far seem to be quite encouraging as regards the economical viability of nuclear desalination options. Thus, for example, specific desalination costs ($/m 3 of desalted water) for nuclear systems, such as the AP-600 and the French PWR-900 (reference base case), coupled to multiple effect distillation (MED) or the reverse osmosis (RO) processes, are 30–60% lower than the desalination costs for fossil energy based systems, using pulverised coal and natural gas with combined cycle, at low discount rates and recommended fossil fuel prices. Even in the most unfavourable scenarios for nuclear energy (discount rate = 10%, low fossil fuel costs) desalination costs with the nuclear reactors are 7–20% lower, depending upon the desalination capacities. Furthermore, with the advanced coupling schemes, utilising waste heat from nuclear reactors, the gains in specific desalination costs of nuclear systems are increased by another 2–15%, even without system and design optimisation. A preliminary evaluation shows that desalination costs with the GT-MHR, coupled to a MED process, could still be much lower than the above nuclear options for desalting capacities ≤ 43 000 m 3 per day. This is because its design intrinsically provides “virtually free” heat at ideal temperatures for desalination (80–100 ◦ C).

In the frame of the future generation of nuclear reactors, liquid metals are foreseen to be used as a primary coolant. Liquid metals are characterized by a very low Prandtl number due to their very high heat diffusivity. As such, they do not meet the so-called Reynolds analogy which assumes a complete similarity between the momentum and the thermal boundary layers via the use of the turbulent Prandtl number. Particularly, in the case of industrial fluid-dynamic calculations where a resolved computation near walls could be extremely time consuming and could need very large computational resources, the use of the classical wall function approach could lead to an inaccurate description of the temperature profile close to the wall. The first aim of the present study is to investigate the ability of a well- established commercial code (ANSYS FLUENT v.14) to deal with this issue, validating a suitable expression for the turbulent Prandtl number. Moreover, a thermal wall-function developed at Universite Catholique de Louvain has been implemented in FLUENT and validated, overcoming the limits of the solver to define it directly. Both the resolved and unresolved approaches have been carried out for a channel flow case and assessed against available direct numerical and large eddy simulations. A comparison between the numerically evaluated Nusselt number and the main correlations available in the literature has been also carried out. Finally, an application of the proposed methodology to a typical sub-channel case has been performed, comparing the results with literature correlations for tube banks.

The project concerns the design, optimization, construction, assembling, start-up and extensive monitoring of an experimental plant consisting of a full-titanium desalinator coupled with a small solar pond. The operational tests took place at the site of the solar pond of the University of Ancona (Italy). Data collected during the start-up and operation of the plant under various conditions are being utilized for improving expertise on heat recovery with highly corrosive fluids, on co-generation plants aimed at producing electricity and fresh water, and on desalination fed by solar energy. Some preliminary economic evaluations are discussed.

The paper deals with the study of an in-vessel LOCA caused by a rupture in the first wall/shielding blanket cooling loop in the ITER plant. The aim is to evaluate the effectiveness of the expansion volume connected to the cooling room and cryopump room in limiting the peak pressure inside the vacuum vessel. The chemical reaction between the coolant and the first wall beryllium layer inside the vacuum vessel has also been considered.

The Power Conversion System (PCS) is designed to remove heat from PHTSs and to ensure an efficient thermal power conversion to electricity. The Dual Coolant Lithium Lead (DCLL) breeding blanket concept is based on two coolant fluids flowing into two independent circuits: a helium circuit, cooling first wall, and a PbLi circuit, removing heat generated in the breeding zone (BZ). The steam generator connected to PbLi system is prototypical, therefore experimental tests are required to demonstrate technological feasibility, performances and to assess capability of codes in providing reliable simulations of the behavior. The objective of the activity is to support the design of a PbLi/water steam generator and the experimental campaign defining test matrix and expected outcomes. Two steam generator configurations are analyzed using a modified version of RELAP5/Mod3.3 (R5): the Steam Generator Bayonet Tube (SGBT) and the Helical Coil Steam Generator (HCSG). The experimental mock-up of these configurations will be tested in CIRCE facility at ENEA Brasimone RC, which is currently the largest and most powerful heavy liquid metal facility in operation. Finally, numerical simulations are employed to support the design of the mock-ups and the experimental campaign.

The water-cooled lithium-lead (PbLi) breeding blanket is one of the candidate systems considered for the implementation in the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This concept employs PbLi liquid metal as tritium breeder and neutron multiplier, water pressurized at 15.5 MPa as the coolant, and EUROFER as the structural material. The current design is based on the single module segment approach and follows the requirements of the DEMO-2015 baseline design. The module is constituted by a basic toroidal-radial cell that is recursively repeated along the poloidal direction where the liquid metal flows along a radial-poloidal path. The heat generated by the fusion reactions is extracted by means of separate cooling systems for the breeding zone and the first wall. A back supporting structure is dedicated to withstand loads of the module during normal and off-normal operations. Water and PbLi manifolds are integrated with primary heat transport and tritium extraction systems. The status of the conceptual design is presented, critically discussing its rationale and main features as supported by neutronic, thermal-hydraulic, magneto-hydrodynamic, and thermo-mechanic analyses. Recent results are outlined, pointing out open issues and development needs.