Experimental and RELAP5-3D results on IELLLO (Integrated European Lead Lithium LOop) operation

The experimental facility IELLLO (Integrated European Lead Lithium LOop) was designed and installed at the ENEA Brasimone Research Centre to support the design of the HCLL TBM (Helium Cooled Lithium Lead Test Blanket Module).This work presents the results of the experimental campaign carried out within the framework of F4E-FPA-372 and which had three main objectives. First, to produce new experimental data for flowing LLE (Lead-Lithium Eutectic) for an analysis of the loop and the characterization of its main components. Then, to evaluate performances of commercial instrumentation as available instrumentation is not designed for use in LLE. Lastly, to use the data for validation of the model developed with the system code RELAP5-3D. The data collected could prove helpful to analyze the behavior of the LLE loop of ITER and DEMO in accidental conditions.The results show that the regenerative countercurrent heat exchanger has an efficiency ranging from 70 to 85%, mainly depending on the LLE mass flow rate. It was verified that the air cooler has the capability to keep the cold part of the loop at 623. K, even in the most demanding situation (700. rpm and maximum temperature of the hot part). The instrumentation tested showed good accuracy, with the exception of the turbine flow meter. Nevertheless, specific limitations in the upper operative temperatures were found for the LLE direct contact pressure transducer. RELAP5-3D simulations fit very well the associated experimental results achieved

Conceptual design of the enhanced coolant purification systems for the European HCLL and HCPB test blanket modules

The Coolant Purification Systems (CPSs) is one of the most relevant ancillary systems of European Helium Cooled Lead Lithium (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Modules (TBMs) which are currently in the preliminary design phase in view of their installation and operation in ITER. The CPS implements mainly two functions: the extraction and concentration of the tritium permeated from the TBM modules into the primary cooling circuit and the chemistry control of helium primary coolant. During the HCLL and HCPB-TBSs (Test Blanket Systems) Conceptual Design Review (CDR) in 2015 it was recognized the need of reducing the tritium permeation into the Port Cell #16 of ITER. To achieve this and, then, to lower the tritium partial pressure in the Helium Cooling Systems in normal operation, the helium flow-rate treated by each CPS has been increased of almost one order of magnitude. In 2017, to satisfy the CDR outcomes and the new design requirements requested by Fusion for Energy (F4E, the European Domestic Agency for ITER), ENEA performed a preliminary design of the “enhanced” CPSs. This paper presents the current design of the “enhanced” CPSs, focusing on design requirements, assumptions, selection of technologies and preliminary components sizing

Experiments on the MHD Effect on the Drainage of a LiPb Channel and Supporting Numerical Computations with the Level Set Method

To analyze the impact of the magnetohydrodynamics (MHD) effect on the fast draining of a LiPb channel (lithium-lead eutectic, 15.7 at. % Li) for a liquid metal fusion blanket such as the water-cooled lithium-lead test blanket system of ITER or DEMO, an experimental campaign was carried out with the support of the Integrated European Lead Lithium LOop experimental facility (IELLLO), installed at the ENEA Brasimone research center, Italy. The experiments were carried out by measuring the drainage time of the internal permanent magnet pump channel, normally used to circulate the LiPb in the loop, with and without the magnetic field. Moreover, this paper proposes a new numerical methodology to study the time delay induced by the MHD by using the commercial software COMSOL Multiphysics. In this way, it was possible to evaluate the LiPb fraction present at each time step in the computational domain and to estimate the time necessary for the complete drainage of the channel. The level set method was used to describe the transient behavior of the MHD flow under low-Rm approximation. The developed code was compared with the experimental results and showed good agreement, and it constitutes the first step in model validation as a possible application to ITER and DEMO. The experimental and numerical analyses performed in this work can be used as a benchmark case for MHD code development

Tritium extraction systems for the European HCLL/HCPB TBMs

One of the most challenging issues for the TBM (Test Blanket Module) testing campaign foreseen in ITER is the operation of TES (Tritium Extraction Systems). This is essential not only to prove the ability to manage correctly the bred tritium but also to validate and qualify the neutronic codes for the prediction of tritium production in view of their use in future fusion plants. Two are the European candidates to be tested in ITER: the HCPB (Helium Cooled Pebble Bed) TBM and the HCLL (Helium Cooled Lithium Lead) TBM. For both these TBM concepts the following points have been addressed in this work: a) the gas stream to be processed by TES b) the TES process flow diagram c) a first assessment of the required spac

Materials selection and design of a hydrogen measurement device in Pb-17Li

In the helium cooled lithium lead (HCLL) and water cooled lithium lead (WCLL) blanket concepts for DEMO correct and reliable management of tritium is of basic importance, both for safety and fuel cycle reasons. To develop a sensor for measurements of hydrogen (and its isotopes) concentration in liquid Pb-17Li, a permeable capsule of niobium was chosen. Different simulations with a mathematical model have been performed, and then the sensor was designed, constructed and tested. The first experimental results in gas phase showed a permeating flux much lower than the predicted one, probably due to the formation of an oxide layer on the capsule surface or to the formation of niobium hydrides. To solve this problem different solutions are presented

A Model for Tritium Transport in Fusion Reactor Components: the FUS-TPC Code

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of atomic hydrogen in materials in which hydrogen and its isotopes are present. In this work the problem of tritium transport from lead-lithium breeder through different heat transfer surfaces to the environment has been studied and analyzed by means of a computational code. The code (FUS-TPC) is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. A simulation, using the code, was performed by adopting the configuration of the European configuration of the Helium Cooled Lead Lithium (HCLL) blanket for DEM