Experimental qualification of new instrumentation for lead-Lithium eutectic in IELLLO facility

The experimental facility IELLLO was installed in ENEA Brasimone R.C. in 2007, aiming to support the design of liquid Test Blanket Modules that will be installed in ITER and to contribute to the development of Lead-Lithium Eutectic (LLE) technologies. IELLLO has been recently upgraded by installing instrumentation relevant for ITER application. Differential pressure transducers, a Coriolis and a thermal mass flow meters were installed in the facility. An experimental campaign was planned, setting two objectives. The first objective was to qualify the instrumentation for flowing LLE The installation of a differential pressure transducer across each flow meter made also possible to characterize the pressure drops across these instruments. The second objective of this activity was to improve the knowledge on the performances of the main components of the loop at lower mass flow rates (namely 0.5-1.2 kg/s) and to quantify their pressure drops. The investigated flow rates were chosen to be relevant for the LLE loop of the WCLL TBS (Water Cooled Lead-Lithium Test Blanket System). This work presents the results of the experimental campaign, paying particular attention to underline the lessons learned on how to correctly operate instrumentation for LLE

GEN-IV LFR development: Status & perspectives

Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of Generation IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to Heavy Liquid Metal (HLM) nuclear reactors. In this frame, ENEA developed one of the larger European experimental fleet of experimental facilities aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for structural materials, and at developing components, instrumentations and innovative systems, supported by experiments and numerical tools. The present work aims at highlighting the capabilities and competencies developed by ENEA so far in the frame of the liquid metal technologies for GEN-IV LFR. In particular, an overview on the ongoing R&D experimental program will be depicted considering the actual fleet of facilities: CIRCE, NACIE-UP, LIFUS5, LECOR and HELENA. CIRCE (CIRColazione Eutettico) is the largest HLM pool facility presently in operation worldwide. Full scale component tests, thermal stratification studies, operational and accidental transients and integral tests for the nuclear safety and SGTR (Steam Generator Tube Rupture) events in a large pool system can be studied. NACIE-UP (NAtural CIrculation Experiment-UPgraded) is a loop with a HLM primary and pressurized water secondary side and a 250 kW power Fuel Pin Simulator working in natural and mixed convection. LIFUS5 (lithium for fusion) is a separated effect facility devoted to the HLM/Water interaction. HELENA (HEavy Liquid metal Experimental loop for advanced Nuclear applications) is a pure lead loop with a mechanical pump for high flow rates experiments. LECOR (LEad CORrosion) is a corrosion loop facility with oxygen control system installed. All the experiment actually ongoing on these facilities are described in the paper, depicting their role in the context of GEN-IV LFR development

Magneto-convective effect on tritium transport at breeder unit level for the WCLL breeding blanket of DEMO

The Water-Cooled Lithium-Lead (WCLL) is one of the four breeding blanket concepts proposed by Europe in view of its DEMO reactor. The velocity field of the electrically conducting lead-lithium eutectic alloy inside the blanket is strongly influenced by the external magnetic field used for plasma confinement combined with buoyancy effect. The strength of the magnetohydrodynamics (MHD) effect and of the magneto-convective effect (MHD and buoyancy force) depends on the intensity of the magnetic field and its orientation with respect to the direction of the lead-lithium motion. This phenomenon significantly influences the resulting temperature and velocity fields, and therefore the tritium transport inside the breeding blanket. A multi-physics approach of a 3D tritium transport model is presented for a simplified geometry of the WCLL breeding blanket. In particular, advection-diffusion of tritium into the lead-lithium eutectic alloy, transfer of tritium from the liquid interface towards the steel, diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant, and advection-diffusion of diatomic tritium into the coolant, temperature field, velocity fields of both lead-lithium and water, buoyancy forces, and MHD effect have been included in this study. The tritium concentrations and the inventories inside the lead-lithium, in the Eurofer pipes and in the baffle, and in the water coolant have been evaluated

Experimental campaign on the upgraded He-FUS3 facility

An extensive thermal-hydraulic experimental campaign was conducted on He-FUS3 helium loop facility to support the conceptual design of HCLL and HCPB Test Blanket System. The experiments were divided into three distinct phases. The first one was dedicated to the evaluation of the new ATEKO Turbo Circulator (TC) performances, identifying its operating limits in terms of supplied helium mass flow as a function of rotational speed, cold by-pass opening and loop pressure. The outcomes were compared with the manufacturer theoretical performance map and with a RELAP5-3D pre-test computation. In the second phase, experiments were carried out to analyze the facility dynamic response in hot conditions and to characterize its main components (TC, heaters, economizer, cooling system and valves). The wide amount of collected data will serve for the development and validation of a numerical model of the facility at TBS conditions. For the third phase, the tests were designed to investigate He-FUS3 behavior in accidental conditions representative of LOFAs and LOCAs scenarios

Design of the Test Section for the Experimental Validation of Antipermeation and Corrosion Barriers for WCLL BB

Tritium permeation into the Primary Heat Transfer System (PHTS) of DEMO and ITER reactors is one of the challenging issues to be solved in order to demonstrate the feasibility of nuclear fusion power plants construction. Several technologies were investigated as antipermeation and corrosion barriers to reduce the tritium permeation flux from the breeder into the PHTS. Within this frame, alumina coating manufactured by Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) are two of the main candidates for the Water Cooled Lithium Lead (WCLL) Breeder Blanket (BB). In order to validate the performance of the coatings on relevant WCLL BB geometries, a mock-up was designed and will be characterized in an experimental facility operating with flowing lithium-lead, called TRIEX-II. The present work aims to illustrate the preliminary engineering design of a WCLL BB mock-up in order to deeply investigate permeation of hydrogen isotopes through PHTS water pipes. The permeation tests are planned in the temperature range between 330 and 500 °C, with hydrogen and deuterium partial pressure in the range of 1–1000 Pa. The hydrogen isotopes transport analysis carried out for the design and integration of the mock-up in TRIEX-II facility is also shown

Characterization of aluminum-based coatings after short term exposure during irradiation campaign in the LVR-15 fission reactor

Protective aluminum-based coatings represent a promising anti-permeation and anti-corrosion barrier for breeding blanket systems developed for European DEMO fusion reactor. Following the prior in-depth characterizations, selected coating candidates were subjected to a combined test consisting of contact with liquid Pb-16Li, its repeated in-situ solidification and re-melting, tritium permeation during gamma and neutron irradiation in the LVR-15 fission reactor. During the sample exposure in the reactor, the temperature of Pb-16Li was between 300 and 425 °C. The irradiation damage averaged over the sample volume estimated by FISPACT code was limited to 0.037 dpa. This article presents post-irradiation characterization of cylindrical Eurofer97 samples coated by electro-chemical X-metal deposition from ionic liquid (ECX) and pulsed laser deposition (PLD) techniques. The coating damage relative to a reference uncoated sample is discussed

Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities

The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021