Cooled Lithium Lead Research Articles

Design of a permeator against vacuum for tritium extraction from eutectic lithium-lead in a DCLL DEMO

One of the most important issues in future fusion power plants is the extraction of tritium generated in the breeders in order to achieve self-sufficiency. When the breeder is a liquid metal one of the most promising techniques is the Permeation Against Vacuum, whose principle is based on tritium diffusion through a permeable membrane in contact with the liquid metal carrier and its further extraction by a vacuum pump. A conceptual design of permeator has been developed, taking into account the features of a DEMO reactor with a Dual Coolant Lithium Lead (DCLL) breeder blanket. The study is based on the analysis of different membranes and geometries aiming at the overall efficiency (extraction capability) of the device, as well as its compatibility with the breeder material. The permeator is based on a rectangular section multi-channel distribution where the liquid metal channels and vacuum channels are alternated in order to maximize the contact area and therefore to promote tritium transport from the bulk to the walls. The resulting permeator design has an excellent estimated extraction efficiency, of 80%, in a relatively compact device.

Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

Quantification of Dominating Factors in Tritium Permeation in PbLi Blankets

In this paper the problem of tritium transport in PbLi (Lead-Lithium) blankets has been studied and analyzed by means of our recently developed computational models. Several simulations are performed by incorporating the geometric configurations of the PbLi blankets including both DCLL (Dual Coolant Lead Lithium) and HCLL (Helium Cooled Lead Lithium) blankets. Tritium permeation loss percentage from the HCLL concept is about one order of magnitude higher than from the DCLL concept (~ 17%. vs. 1.2%). Sensitivity study also shows that the most relevant factors on tritium permeation are: 1) the level of tritium solubility in PbLi, 2) the gap velocity of the liquid metal in a DCLL blanket, 3) Hartmann number, and 4) the FCI (Flow Channel Insert) electrical conductivity.

Thermo-fluid dynamics and corrosion analysis of a self cooled lead lithium blanket for the HiPER reactor

The HiPER reactor is the HiPER project phase devoted to power production. To reach a preliminary reactor design, tritium breeding schemes need to be adapted to the HiPER project technologies selection: direct drive ignition, 150 Hz of power released through fusion reactions, and the dry first wall scheme. In this paper we address the main challenge of the HiPER EUROFER-based self cooled lead lithium blanket, which is related to the corrosive behavior of Pb–15.7Li in contact with EUROFER. We evaluate the cooling and corrosion behavior of the so-called separated first wall blanket (SFWB) configuration by performing thermo-fluid dynamics simulations using a large eddy simulation approach. Despite the expected improvement over the integrated first wall blanket, we still find an unsatisfactory cooling performance, expressed as a low outlet Pb–15.7Li temperature plus too high corrosion rates derived from local Pb–15.7Li high temperature and velocity, which can mainly be attributed to the geometry of the channels. Nevertheless, the analysis allowed us to devise future modifications of the SFWB to overcome the limitations found with the present design.

Tritium transport calculations for the IFMIF Tritium Release Test Module

The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the computations for the five computational cases concerning safety and experiment analyses are presented.

Progress in EU Blanket Technology

In the last years EU started a new effort for the definition of a Roadmap for the fusion beyond ITER up to the construction of the DEMO reactor. The blanket technology, key issue for the development of the fusion breeder reactor, is an important part of this review. In the last 10 years the R&D of EU in blanket has been focused on two kinds of blanket concepts, namely the Helium Cooled Pebble Bed (HCPB) and Lithium lead (HCLL); both are EU reference concepts for DEMO and for the test in ITER. In addition to these two reference concepts, few other concepts have been investigated in EU. One of these is the Dual Coolant Lithium Lead (DCLL) that was developed in the 90-ties in Karlsruhe and UCSD. The spectrum of blankets studied in EU includes also a water cooled blanket with liquid breeder, the WCLL.

Sensitivity Study for Tritium Permeation in Helium-Cooled Lead-Lithium DEMO Blanket with 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 hydrogen and its isotopes in materials. We address the problem of tritium transport in Helium Cooled Lead-Lithium (HCLL) DEMO blanket from lead-lithium breeder through different heat transfer surfaces to the environment by developing a computational code (FUS-TPC). The main features of the code are briefly described and a parametric study is performed in order to identify the most influencing parameters in terms of tritium releases into the environment and of tritium inventories. The results showed that the results are strongly affected by the tritium Sievert’s constant in Lead-Lithium and the efficiency of permeation barriers.

Tritium transport analysis in HCPB DEMO blanket with the FUS-TPC code

In the breeding areas a significant fraction of the tritium produced is extracted out from the Breeding Zone by the He gas purging the breeding ceramic in the helium cooled pebble bed (HCPB) blanket concept or transported in solution by the flowing liquid metal alloy in the helium cooled lead lithium (HCLL) blanket concept and then extracted outside the blanket box. Tritium produced inside the breeder can enter the metal structures and can be lost by permeation to the environment. It should therefore be kept under close control throughout the fusion reactor lifetime.In this study the problem of tritium transport in HCPB DEMO blanket has been studied and analysed by means of the computational code FUS-TPC, which has been originally developed to study the tritium transport in HCLL blanket and it has the main goal to provide the total tritium losses into the environment and the tritium inventories inside several HCPB blanket locations (e.g. purge gas and main coolant loops).

Magneto-convective flows in electrically and thermally coupled channels

A liquid metal blanket concept using a PbLi alloy as breeder material and helium as coolant is one of the proposed designs that will be tested in the experimental fusion reactor ITER. In order to finalize the design of the helium cooled lead lithium (HCLL) blanket, studies are still required to fully understand the behavior of the electrically conducting breeder under the influence of the intense magnetic field that confines the fusion plasma. Liquid metal HCLL blanket flows are expected to be mainly driven by buoyancy forces caused by non-isothermal operating conditions due to neutron volumetric heating, since only a weak forced flow is required for tritium extraction. The present numerical study aims at clarifying the influence of electromagnetic and thermal coupling of neighboring fluid domains on magneto-convective flows in geometries relevant for the HCLL blanket concept. Velocity and temperature distributions are discussed for various volumetric heat sources.

Studies of a self-cooled lead lithium blanket for HiPER reactor

Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall – Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates.

Design of a multipurpose laboratory scale apparatus for the investigation of hydrogen isotopes in PbLi and permeation technologies

The knowledge of the tritium transport parameters in lead lithium is fundamental for the design of the HCLL (helium cooled lead lithium) blanket. In fact, the inventory of tritium in fusion reactors blankets and the permeation of tritium into the blanket coolant, with the consequent leaks toward the environment, are strongly depending on its solubility and diffusivity in the lead alloy PbLi. Several experiments, devoted to investigate the function linking the tritium solubilised in lead lithium with the corresponding tritium partial pressure at equilibrium, were carried out in the past, but significant uncertainties still remain.A detailed analysis of the past experimental works is carried out in this paper with the aim to investigate the main problems occurred in the facilities used to measure the tritium solubility in PbLi that caused such a big spread in the achieved results. On the basis of this analysis, a new a multipurpose laboratory scale apparatus has been designed. The apparatus is able to measure the tritium solubility and diffusivity in PbLi in the range of temperature 300–550°C and it will be operated with hydrogen partial pressure in the range 102–104Pa. The facility can work with desorption and absorption technique.Moreover, the apparatus has been designed to allow the testing of H/D concentration sensors in Pb–15.7Li in operative conditions relevant to the HCLL–TBM and the characterisation of hydrogen permeation barrier.

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 DEMO.

Thermal Insulator of Porous SiC/SiC Composites for Fusion Blanket System

For fusion energy realization, attractive energy conversion system and fuel supply system are essentially required. The current ITER test blanket module study has dual coolant lithium lead (DCLL) system as the attractive blanket option. The concept of making ceramic thermal insulation panels for DCLL blanket has been presented, however, realistic material systems have not been proven so far. This work is trying to present the porous SiC/SiC thermal insulation panel based on the current NITE-method technology. Typical panels with through thickness and in-plane channels and porosities tailored were designed, fabricated and evaluated, successfully. Thermal conductivity of those panels were controlled, as designed, and the tailoring capability of thermal conductivity by NITE-method was presented. This concept and technology can satisfy the basic material requirements as well as economical requirements and large scale production requirements.

Development of a Numerical Tool to Simulate Magnetohydrodynamic Interactions of Liquid Metals with Strong Applied Magnetic Fields

In the framework of the study of a European helium cooled lead lithium blanket concept for ITER, numerical tools are developed to complement experimental activities. Full capability to simulate numerically the global magnetohydrodynamic flow and pressure distributions resulting from the interaction of the liquid metal with the strong plasma confining magnetic field is not achieved yet. Calculations should support the selection and validation of physical models for 3D coupled phenomena, like magneto-convection, as well as for corrosion and tritium permeation processes. Moreover, simulations help to interpret measurement data and to enhance the development of extrapolation procedures from small-scale experiments to a DEMO reactor.The present paper summarizes the mathematical algorithm and modeling requirements for accurate predictions of liquid-metal flows under very intense magnetic fields in geometries with arbitrary electric conductivity of the walls. The Lorentz force term and additional equations determining electric current density and potential have been introduced in a consistent and conservative way into the existing hydrodynamic open source code OpenFOAM. The use of non-orthogonal corrections leads to a significant improvement of the MHD code at fusion relevant strong magnetic fields. The discussion focuses on benchmark problems used to validate the new developed tool and on the treatment in OpenFOAM of MHD flows in geometries with walls of finite electric conductivity. According to the authors’ knowledge, the implementation of this capability in this open source code has not been reported so far in other references.

Effects of Radial Variation of the Magnetic Field on the Pressure Distribution in the European Liquid-Metal Blanket Concept

Liquid-metal flows in the European helium cooled lead lithium blanket are strongly affected by the intensity and the distribution of the externally applied magnetic field required for plasma confinement. An experimental campaign has been performed to investigate the pressure distribution of magnetohydrodynamic flows in a scaled model of a liquid-metal blanket module.A variety of experiments has been carried out to asses the influence of flow rates and of the strength and non-uniformity of the magnetic field on the pressure distribution in the test-section. The magnetic field available in the laboratory is characterized by a large zone of uniform magnitude and gradients at the entrance and the exit. The mock-up has been located at various positions along the magnet axis to reproduce operating conditions in which the toroidal field varies in radial direction, i.e. it changes from the back plate to the first wall. Measurements show that the magnitude of the total pressure drop in the mock-up is significantly influenced by the strength of the local magnetic field at the manifolds, while gradients across the breeder units have minor effects. This study confirms the critical role of manifolds in determining the total pressure drop in the blanket.

Magnetohydrodynamic pressure drops in geometric elements forming a HCLL blanket mock-up

The helium cooled lead lithium (HCLL) blanket concept is one of the designs that will be tested and validated in the experimental reactor ITER. The liquid metal, which is slowly circulated in the blanket, interacts with the magnetic field that confines the fusion plasma leading to modifications of velocity and pressure distributions compared to the ones in hydrodynamic flows. Experiments have been performed to investigate magnetohydrodynamic flows in a scaled mock-up of a HCLL blanket. One of the aims is identifying the geometric elements that give rise to the main contributions to the total pressure head in the module. Parametric studies have been carried out to assess the influence of flow rate and magnetic field intensity on pressure drops.

Analysis of in-pile operation and modeling of tritium transport in lead-lithium irradiation experiments Libretto 4/1, 4/2 and 5

In the framework of the European R&D for the Helium Cooled Lead Lithium (HCLL) test blanket concept for ITER, NRG has designed three new Libretto irradiation experiments. Two experiments, Libretto 4/1 and 4/2, have been irradiated in HFR for 730 Full Power Days. The objective was to monitor on line the in-pile tritium production and permeation through Eurofer in contact with lead lithium eutectic. The irradiation of Libretto 4/1 and 4/2 has been conducted at 550 °C and 350 °C, respectively. Libretto 5 experiment has been irradiated for 520 Full Power Days in the temperature range of 300–500 °C with the objective to study tritium behavior in a lead lithium eutectic. The operation of the Libretto experiments included in-pile variations of hydrogen concentration in the purge gas and temperature transients. The irradiation of all three Libretto experiments has finished in 2009 and this work reports the analysis of the in-pile data. To analyze the data a model of tritium transport in lead lithium has been developed. The model incorporates two transport mechanisms: bulk diffusion and via rising tritium-helium bubbles. It was demonstrated that the bubble transport mechanism is much faster and in better agreement with the tritium residence times measured in the Libretto experiments.

Influence of helium cooling channels on magnetohydrodynamic flows in the HCLL blanket

One of the blanket concepts proposed to be tested in ITER as part of the test blanket module program of the European Union is the helium cooled lead lithium blanket design. In this configuration the so called breeder units are arranged in an array, separated by a stiffening grid, to form blanket modules. The deposited thermal energy is removed by helium flowing at high pressure and speed in channels integrated both in the walls and in cooling plates that subdivide the breeder units into flat ducts where the lead lithium circulates under the influence of the strong plasma confining magnetic field. This gives rise to magnetohydrodynamic (MHD) phenomena whose effects on flow distribution have to be investigated to evaluate the performance of the proposed design. The established MHD flow is affected by the presence of helium channels in cooling and stiffening plates that results in non-homogeneous wall conductance. In support to the conceptual study of a liquid metal blanket, numerical investigations of fully developed MHD flows in a central cross-section of breeder units have been performed, taking into account both the presence of helium channels in the walls and the multi-channel effects caused by the exchange of currents through walls separating different fluid domains.

Experimental study of efficiency of natural oxide layers for reduction of tritium permeation through Eurofer 97

Tritium permeation from the breeder through the helium coolant is a fundamental safety issue in the design of the HCLL (Helium coolant lithium lead) blanket system. The permeation of hydrogen isotopes through Eurofer in different conditions was deeply studied in the past, demonstrating that it is necessary to reduce this amount using tritium permeation barriers (TPB). A strong effort has been made to select the best technological solution for the realisation of tritium permeation barriers on complex structures not directly accessible after the completion of the manufacturing process, but after many years of activity for the qualification of different materials acting as TPBs it was demonstrated that these technologies are not yet mature for nuclear applications. An easier solution was identified in the nucleation and growth of natural oxides on the helium-exposed surface of cooling system components. The major objective of this work is the evaluation of the Permeation Reduction Factor (PRF) of natural oxides on Eurofer steel adding a known content of water and hydrogen to argon, used in substitution of helium. The PRF was measured on disk shaped specimens, in gas phase, using the PERI II apparatus, at a temperature of 550 °C. The oxide layer was produced in situ, in a well-defined range of hydrogen on water ratios. The obtained results are presented and discussed.

Electric flow coupling in the HCLL blanket concept

A modular helium cooled lead lithium (HCLL) blanket concept, in which helium is used to cool the breeder zone through cooling plates immersed in the liquid metal as well as the whole structure (first wall and stiffening plates), has been selected as part of the test blanket module (TBM) program of the European Union. In this blanket concept there are issues related to the magnetohydrodynamic (MHD) interaction of the moving liquid metal with the magnetic field that have to be addressed in order to assess the feasibility of this blanket design. Since the walls and the cooling plates are electrically conducting, an exchange of electric currents may occur that leads to an electric coupling of the flow in adjacent fluid domains. Fully developed MHD flows have been investigated numerically in four breeder units containing cooling plates. Results obtained for a pure toroidal magnetic field and for fields having a certain inclination with respect to the toroidal direction are summarized in this paper focusing on current and velocity distribution.