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The present paper provides a discussion of the thermal-hydraulics requirements in fusion reactor components, with particular reference to the removal of high heat flux from plasma facing components and the critical heat flux (CHF) limit. Available experimental data on CHF of subcooled flow boiling in water, in the ranges of interest of fusion reactors thermal-hydraulic conditions, i.e. high inlet subcooling and velocity, and small channel diameter and length, are analyzed to discuss the influence of these parameters on CHF. The reference data-set (1887 experimental points) covers a wide range of operating conditions in the frame of present interest (0.1 < p < 8.4 MPa; 0.3 < D < 25.4 mm; 0.25 < L < 61 cm; 900 < G < 90000 kg/m2·s; 0.3 < Tin < 242.7°C). The aim of the research was to identify a new correlation based on a structure representing the relation of heat balance and using a non-linear regression analysis of the available data-set. A preliminary correlation (DINCE-92), based on 544 data points, had been developed providing a sensible improvement in predictions with respect to available predictive tools. Now, a new correlation (DINCE-93), based on the same structure of the above one and characterized by a very good statistics using a total of 1887 experimental points (88% of predictions are within ±20%) and by an R.M.S. error of 14.2%, has been identified and analyzed.

The interaction of the liquid metal with the plasma confinement magnetic field constitutes a challenge for the design of fusion reactor blankets, due to the arise of MHD effects: increased pressure drops, heat transfer suppression, etc. To overcome these issues, a dielectric fluid can be employed as coolant for the breeding zone. A typical configuration involves pipes transverse to the liquid metal flow direction. This numerical study is conducted to assess the influence of pipe conductivity on the MHD flow and heat transfer. The CFD code ANSYS CFX was employed for this purpose. The fluid is assumed to be bounded by rectangular walls with non-uniform thickness and subject to a skewed magnetic field with the main component aligned with the cylinder axis. The simulations were restricted to Re = (20; 40) and M = (10; 50). Three different scenarios for the obstacle were considered: perfectly insulating, finite conductivity and perfectly conducting. The electrical conductivity was found to affect the channel pressure penalty due to the obstacle insertion only for M = 10 and just for the two limiting cases. A general increment of the heat transfer with M was found due to the tendency of the magnetic field to equalize the flow rate between the sub-channels individuated by the pipe. The best results were obtained with the insulating pipe, due to the reduced electromagnetic drag. The generation of counter-rotating vortices close to the lateral duct walls was observed for M = 50 and perfectly conducting pipe as a result of the modified currents distribution.

Computer modeling codes represent a crucial tool to support the design of future fusion plants. Since one of the most important functions that codes must achieve is assessing safety systems design, a verification and validation phase is required. In this framework, a code-to-code comparison among four codes has been carried within EUROfusion consortium between Safety Analyses and Environment and Balance of Plant work packages. In particular, an in-vessel Loss of Coolant Accident has been selected as a benchmark scenario for investigating thermal-hydraulic parameters of the vacuum vessel and its suppression system. This paper aims to compare the answer of different codes in terms of peak pressure within the vacuum vessel and its timing, the equilibrium pressure, and relief valve and rupture disk opening using a simplified thermal-hydraulic model of the EU-DEMO. In order to minimize the differences among the codes, the models have been kept to the simplest possible nodalization, successively increasing the model complexity.

The Water-Cooled Lithium Lead (WCLL) is a blanket concept pursued in the framework of Test Blanket Module (TBM) campaign in ITER. Even if the liquid metal is circulated slowly in the component, magnetohydrodynamic (MHD) pressure losses are still expected to be significant. The aim of this paper is to assess the MHD pressure losses in the TBM frontal part, also called Breeding Zone (BZ). There, important contributions are caused by the manifold interface, the presence of cooling pipes obstructing the fluid movement, a sharp hairpin bend, and non-uniform wall thickness of the walls. Direct numerical simulation of 2D and 3D MHD flows is used to estimate the head loss for each one of these elements. A scaling law is derived to allow quick estimate of the pressure loss from reference parameters. The main contribution to the head loss is caused by the windows that connect the BZ with the manifold.

CIRCE is a lead–bismuth eutectic alloy (LBE) pool facility aimed to simulate the primary system of a heavy liquid metal (HLM) cooled pool-type fast reactor. The experimental facility was implemented with a new test section, called HERO (Heavy liquid mEtal pRessurized water cOoled tubes), which consists of a steam generator composed of seven double-wall bayonet tubes (DWBT) with an active length of six meters. The experimental campaign aims to investigate HERO behavior, which is representative of the tubes that will compose ALFRED SG. In the framework of the Horizon 2020 SESAME project, a transient test was selected for the realization of a validation benchmark. The test consists of a protected loss of flow accident (PLOFA) simulating the shutdown of primary pumps, the reactor scram and the activation of the DHR system. A RELAP5-3D© nodalization scheme was developed in the pre-test phase at DIAEE of “Sapienza” University of Rome, providing useful information to the experimentalists. The model consisted to a mono-dimensional scheme of the primary flow path and the SG secondary side, and a multi-dimensional component simulating the large LBE pool. The analysis of experimental data, provided by ENEA, has suggested to improve the thermal–hydraulic model with a more detailed nodalization scheme of the secondary loop, looking to reproduce the asymmetries observed on the DWBTs operation. The paper summarizes the post-test activity performed in the frame of the H2020 SESAME project as a contribution of the benchmark activity, highlighting a global agreement between simulations and experiment for all the primary circuit physical quantities monitored. Then, the attention is focused on the secondary system operation, where uncertainties related to the boundary conditions affect the computational results.

The Water Cooled Lithium Lead (WCLL) is one of the selected breeding blanket (BB) concepts to be investigated in the EUROfusion Breeding Blanket Project (WPBB), and it was also recently chosen as one of the mock-up for ITER Test Blanket Module (TBM) program. The program foresees the test of different BB mock-ups, called Test Blanket Modules, with all the related ancillary systems. A pre-conceptual design of the Water Cooling System (WCS) of the ITER WCLL-TBM was developed considering the same cooling function of the EU-DEMO WCLL-BB primary heat transfer system (PHTS), but matching different boundary conditions: a scaled source power and far lower heat sink temperatures. A complete thermal-hydraulic (TH) model of the WCS loop and TBM set was developed using a modified version of RELAP5/Mod3.3 system code to verify component sizing and to investigate the system behavior during steady-state and transient conditions. The full plasma power scenario was simulated and used as an initial condition for transient calculations. ITER Normal Operational State (NOS) was studied to evaluate the system response. Simulation results highlighted the need for an electric heater to keep the WCS system in stable operation. A sensitivity analysis was carried out to optimize the heater duty cycle.

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 liquid metal is distributed to and collected from the breeding zone through a compact poloidal manifold composed of two co-axial rectangular channels. The external channel, tasked with distribution, and the internal one, assigned to the collection, are co-flowing and share an electrically conductive wall ( c w = 0.1 ). The liquid metal, interacting with the reactor magnetic field, leads to the arising of MagnetoHydroDynamic (MHD) effects that are expected to significantly modify the flow feature and electrically couple the external and internal channels. In this work, the general-purpose CFD code Ansys CFX 18.2 is used to study the coupling phenomena in a wide range of magnetic fields (up to H a = 2000 ) for a prototypical square co-axial channel. Characteristic flow features and their evolution with increasing magnetic field and varying mass flow rate between the channels are discussed and compared with the uncoupled case, which is in turn composed by a rectangular electro-conductive annulus (external) and a square electro-conductive duct (internal). A correlation is found linking the pressure loss in the studied configuration and an equivalent square channel through a corrective factor e, which exhibits an asymptotic behavior for H a > 1000 .

The way to arrive at a licensing phase for a nuclear fusion installation is not straightforward mainly because of the lack of operating experience and of dedicated nuclear regulations. In fact, only small/medium experimental facilities exist with limited licensing processes and only one large experiment, ITER, has obtained a construction license. Therefore, the safety assessment and the preparation of the preliminary safety report is almost a first of a kind for DEMO. Taking advantage of the fission power plants experience and considering to the maximum extent the ITER safety studies, the preparation of a Generic Site Safety Report (GSSR) has begun. It will require some years to be completed; however currently, at the starting point, the strategy to develop it is clear and well defined. This paper considers all the safety issues that will be included in the GSSR because they have been clarified in the frame of the European Workprogramme for DEMO from 2014 up to 2018, and should not be modified in the future, such as the safety requirements for the plant and the systems, the tools to be used for the safety assessment, the procedures for the selection of the reference accidents, and so on. Together with these topics, considered as goals achieved, there are others for which an additional effort is necessary because they do not cover all the expected requirements of the likely licensing procedures applicable for DEMO. A complete spectrum of Design Basis Accidents and Beyond Design Basis Accidents that can determine the risk of releases from the main systems of the power plant is still incomplete together with the safety classification of most of the Structures, Systems and Components, and the feasibility and analyses of some accident mitigation systems. The outcome of this study is the quantification, when possible, of the gap between the results achieved and the goals established in the power plant guidelines. It will help also to qualify the effort required in terms of studies, experiments and human resources to reach a good stage for successful DEMO licensing.