Helium Cooled Lithium Lead Blanket Research Articles

Experimental investigation of MHD flows in a WCLL TBM mock-up

Experiments have been performed in the MEKKA laboratory of the Karlsruhe Institute of Technology to characterize the influence of a magnetic field on liquid metal flows in a scaled mock-up of the water-cooled lead-lithium test blanket module of ITER. The test section consists of eight breeder units that are fed and drained by long electrically coupled manifolds oriented along the poloidal direction. Pressure differences between several points of the mock-up have been recorded for various liquid metal flow rates and strengths of the imposed magnetic field. The main contributions to the total pressure drop in the test section have been identified as a function of characteristic flow parameters. For sufficiently strong magnetic fields, the experimental data shows that the non-dimensional pressure loss is practically independent of the flow rate, i.e. inertia forces become negligible. The experiments also confirm previous conclusions of magnetohydrodynamic experiments performed in a scaled-down mock-up of a helium-cooled lead-lithium blanket, namely that the largest pressure drop in the blanket module originates from the flow in the distributing and collecting manifolds. Moreover, the experimental study demonstrates that the current manifold design does not allow the flow to be uniformly distributed among all the breeder units.

Pressure wave behavior and its effects on structure under In-box LOCA in a helium-cooled lead lithium blanket of hydrogen fusion reactors

The helium-cooled lead lithium (PbLi) blanket is considered as one of the candidate blanket concepts selected for the hydrogen fusion DEMO reactors and beyond, which has the advantages of simple structure, strong heat removal capacity and high tritium breeding ratio. However, due to the harsh environment such as high-energy neutron irradiation, high thermal load and great pressure gradient, there is a high possibility that one or some of the thousands of coolant channels will break in the breeding zone, which is so-called In-box Loss of Coolant Accident (In-box LOCA). When the accident occurs, the high pressure helium will rapidly inject into the lead lithium flow channel, generating a complex two-phase flow and great pressure shock effect, which may cause the peak pressure to exceed the design limit and threaten the integrity of the blanket structure. Therefore, it is of great significance to perform the transient analysis of in-box LOCA to improve the safety of the blanket and avoid the leakage of radioactive materials. In this paper, a two-way coupling model for fluid-solid interaction was established based on the ANSYS Workbench, and the model were validated through the experimental data obtained by injecting the high pressure helium gas into liquid lithium lead. Then the validated model was applied to the transient pressure wave propagation analysis and structural stress analysis of the Dual-Functional Lithium Lead (DFLL) blanket in order to explore the integrity of blanket structure under In-box LOCA. In addition, the effects of break location on pressure and structural stress was also investigated through six cases. The study found that the transient pressure in the DFLL blanket gone through three stages in any case: step rise, oscillate, and flatten out. Pressure peaks occurred during oscillations and their values were strongly dependent on the break location. The closer to the inlet/outlet, the higher the peak pressure was. The maximum pressure reached more than twice of the inlet pressure (up to ~16 MPa). As a result, the structural stress in some local areas has exceeded the allowable limits, and the corresponding suggestions for improvement have also been put forward. This study can provide guidance for safety design, operation and accident mitigation measures of helium-cooled lead lithium blankets.

Studied of Tritium Breeding Ratio on HCLL (Helium Cooled Lithium-Lead) Fusion Blanket by Using MCNP Program

The fusion reactor using a tritium-deuterium reaction is going to be a prolific source of 14.1 MeV (80% of fusion energy) neutrons, which will release from the plasma. This reactor features a major radius of 8.1 m and a minor radius of 2.5 m. Thus, one part of the fusion reactor, which is a basic design is Blanket Module (BM). One of the two European blanket concepts is Helium Cooled Lithium Lead (HCLL). HCLL blanket uses Pb-Li as a neutron multiplier, Li as a breeder, and He as a cooler is in high pressure (8 MPa). The study goal is to investigate the feasibility of blanket reactors, such as neutron analysis, shield efficiency, and tritium breeding capabilities. This calculation was simulated to achieve TBR value more than one, which will assure tritium self-sufficiency in steady-state operations. The MCNP program with Monte Carlo is a tool for calculating this. The neutron flux produced by this simulation shows that the reactor does no leak. And the production of tritium from breeding blanket will be optimal when the utilizing of breeding materials is Pb-Li.

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.

Measurements of neutron transport of well defined silicon filtered beam in lead

The correct description of neutron transport in lead is an essential task for correct description of tritium production in the DEMO (DEMOnstration Power Station) breeding blanket because some concepts deal with lead as a major component: namely the WCLL (water cooled lithium lead blanket), HCLL (helium cooled lithium lead blanket), and DCLL (dual cooled lithium lead blanket). Concerning the improvement of the knowledge about the transport of fast neutrons in lead, a set of experiments and calculations was carried out to study this problem with a well-defined neutron beam. The neutron flux behind various lead arrangements positioned along the beam axis was measured using a stilbene scintillation crystal (10 mm × 10 mm) with neutron and gamma pulse shape discrimination. The measurement was performed along the beam axis and in the case of the thick target also above the axis, to estimate the neutron angular scatter in lead. The calculations were realized using MCNP6 with various nuclear data libraries. Discrepancies in the angular distribution description in the energy region of about 1 MeV were discovered by these experiments.

Tritium transport model at the minimal functional unit level for HCLL and WCLL breeding blankets of DEMO

The Helium-Cooled Lithium-Lead (HCLL) and Water-Cooled Lithium-Lead (WCLL) Breeding Blankets are two of the four European blanket designs proposed for the DEMO reactor. A tritium transport model inside the blanket is fundamental to assess its preliminary design and safety features. Tritium transport and permeation are complex phenomena to be taken into account in the evaluation of the tritium balance, in order to guarantee tritium self-sufficiency and to characterise tritium concentrations, inventories and losses. In this context, the study has been performed at the minimal functional unit level of the outboard equatorial breeding blanket module, which is, during continuous operative conditions, one of the most loaded modules and this results in higher permeation phenomena. For these purposes, a 2D model for the breeder unit of HCLL and a 3D model for the single cell of WCLL breeding blanket concepts have been investigated. The models include 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. Thermal field, tritium generation rate profile, velocity field of lead-lithium and coolant have been also taken into account.

Determination of multichannel MHD velocity profiles from wall-potential measurements and numerical simulations

In the helium cooled lead lithium (HCLL) blanket concept the eutectic alloy PbLi is used as breeder material and the heat is removed by helium. The liquid metal flows in rectangular boxes, called breeder units (BUs), which are arranged in columns to form a blanket module. Experiments have been performed to study MHD flows in a HCLL blanket mock-up that consists of 8 BUs fed and drained by poloidal manifolds. Electromagnetic flow coupling of adjacent fluid domains, due to an exchange of electric currents, plays an essential role in determining the velocity distribution in the breeding zone. During experiments electric potential on the walls perpendicular to the magnetic field has been recorded. In general, the electric potential serves as approximate streamfunction of the flow if walls are thin and induced currents are small. However, if the walls of the mock-up are thick, as in the performed experiments, induced currents should be taken into account. In order to correctly interpret surface electric potential measurements in terms of velocity distribution inside the test-section, a comparison with numerical simulations is required. A procedure has been developed to derive global solutions for magnetohydrodynamic (MHD) flows and to predict the velocity distribution in the test-section starting from experimental potential measurements and by comparing them with numerical results.

Experimental investigation of MHD pressure losses in a mock-up of a liquid metal blanket

Experiments have been performed to investigate the influence of a magnetic field on liquid metal flows in a scaled mock-up of a helium cooled lead lithium (HCLL) blanket. During the experiments pressure differences between points on the mock-up have been recorded for various values of flow rate and magnitude of the imposed magnetic field. The main contributions to the total pressure drop in the test-section have been identified as a function of characteristic flow parameters. For sufficiently strong magnetic fields the non-dimensional pressure losses are practically independent on the flow rate, namely inertia forces become negligible. Previous experiments on MHD flows in a simplified test-section for a HCLL blanket showed that the main contributions to the total pressure drop in a blanket module originate from the flow in the distributing and collecting manifolds. The new experiments confirm that the largest pressure drops occur along manifolds and near the first wall of the blanket module, where the liquid metal passes through small openings in the stiffening plates separating two breeder units. Moreover, the experimental data shows that with the present manifold design the flow does not distribute homogeneously among the 8 stacked boxes that form the breeding zone.

Pressure distribution in MHD flows in an experimental test-section for a HCLL blanket

Liquid metal magnetohydrodynamic (MHD) flows in a scaled mock-up of a Helium cooled lead lithium blanket have been investigated experimentally in the MEKKA laboratory at the Karlsruhe Institute of Technology. The test section consists of 8 breeder units (BUs) connected two by two at the first wall (FW) through a series of small openings. The large boxes are fed and drained by poloidal manifolds formed by two parallel channels separated by a step-shaped wall, which is electrically conducting. The complexity of the geometry and the strong 3D MHD effects that characterize the flow in the mock-up are at the present time still very challenging for numerical simulations. Therefore, experiments are required to obtain an overview of pressure and velocity distributions in such geometry and to derive scaling laws to extrapolate the results to realistic ITER operating conditions. Pressure differences between several points on the test section have been measured. The influence of magnetic field strength and flow rate on pressure losses in the mock-up has been studied. The largest pressure drops occur along the manifolds. Another significant part of the entire pressure drop comes from flow near the FW, where the liquid metal passes through small gaps into the adjacent BU. This pressure drop between external pairs of BUs is at least 4–5 times larger than that in the internal ones. This suggests an inhomogeneous partitioning of the liquid metal flow among the BUs, which is related to the geometry of the manifold.

Quantifying TBR uncertainty due to lead nuclear data in HCLL blanket modelling by the Total Monte-Carlo method

The design of complex and expensive fusion breeder blanket assemblies requires detailed engineering studies that rely upon nuclear analyses to determine essential operating parameters, including tritium production rate, nuclear heating and radiation damage. Lack of precise knowledge of input nuclear physics, referred to as nuclear data, results in simulation uncertainties to which some parameters are highly sensitive. In this work, a set of Pb TENDL-2015 neutron-induced files generated using sampled input parameters are employed using the Total Monte-Carlo method to quantify the uncertainty of tritium breeding in a Helium Cooled Lithium Lead blanket of a demonstration fusion plant. Fully correlated uncertainties including neutron multiplicity reactions and elastic scattering result in a tritium production uncertainty of ±1.2%(1σ).

Activation and decay heat analysis of the European DEMO blanket concepts

Demonstrating tritium self-sufficiency is an important goal of the European tokamak demonstration fusion reactor. Currently four breeder blanket concepts are being considered: the Helium Cooled Pebble Bed (HCPB), Helium Cooled Lithium-Lead (HCLL), Dual Cooled Lithium-Lead (DCLL) and Water Cooled Lithium-Lead (WCLL). Differences in materials and construction of the four breeder blanket concepts lead to differing nuclear responses. As well as affecting tritium breeding this is also of particular importance in safety analyses, such as the modelling of loss of coolant accidents, as it affects the blanket's decay heat and nuclide inventory.This paper presents and discusses analysis performed for each of the 2014 designs of the blanket concepts to ascertain the decay heat and nuclide inventory for the entire reactor. It was found that the total decay heat at short decay times for the HCLL concept (17.5MW at 1s) was between 17% and 22% lower than the HCPB, WCLL and DCLL. At longer decay times (∼100 years) it was found that the DCLL and WCLL blankets had decay heats in the region of 2–3 orders of magnitude above the HCPB and HCLL blankets. The differences noted between the blanket concepts are discussed in terms of neutron spectrum and material composition.

Influence of modifications of HCLL blanket design on MHD pressure losses

In 2008–2009 experiments have been performed to investigate liquid metal magnetohydrodynamic (MHD) flows in a scaled mock-up of a helium cooled lead lithium (HCLL) blanket. In order to improve the mechanical stiffness of HCLL blanket modules the design of the stiffening plate between two hydraulically connected breeder units (BUs) has been later modified. In the former design the liquid metal passed from one BU into the adjacent one by flowing through a narrow gap that extended along the entire width of the BU. In the most recent design this opening has been replaced by a series of smaller gaps. Therefore the velocity increases locally owing to the reduced cross-section along the flow path and the liquid metal contracts and expands along magnetic field lines to enter the neighboring BU. These flow conditions are known to create additional MHD pressure losses. In order to estimate the influence of design changes on MHD flow and pressure drop the previous test section has been adapted to the new design. Experimental results show that the modifications of the design at the first wall lead to an increase of pressure drop near the first wall by a factor 3–3.5. As a consequence the total pressure drop becomes larger. The 3D MHD phenomena that occur at the first wall seem to be mainly related to inertia effects that are confined in layers parallel to the magnetic field.

A Methodology for Accident Analysis of Fusion Breeder Blankets and Its Application to Helium-Cooled Lead–Lithium Blanket

“Fusion for Energy” (F4E) is designing, developing, and implementing the European helium-cooled lead–lithium (HCLL) and helium-cooled pebble-bed test blanket systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and Idaho National Laboratory efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to the ITER accident analyses. The methodology phases are illustrated in this paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

Tritium transport in HCLL and WCLL DEMO blankets

The Helium-Cooled Lithium Lead (HCLL) and Water-Cooled Lithium Lead (WCLL) Breeding Blankets are two of the four blanket designs proposed for DEMO reactor. The study of tritium transport inside the blankets is fundamental to assess their preliminary design and safety features. A mathematical model has been derived, in a new form making makes easier to determine the most critical components as far as tritium losses and tritium inventories are concerned, and to model the tritium performance of the whole system. Two cases have been studied, the former with tritium generation rate constant in time and the latter considering a typical pulsed operation for a time span of 100h. Tritium inventories and tritium losses are the main output of the model. Tritium concentrations, inventories and losses are initially calculated and compared for the two blankets, in a reference case without permeation barriers or cold traps. A parametric study to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and/or losses, has been carried out.

MHD PbLi experiments in MaPLE loop at UCLA

Experiments on magnetohydrodynamic (MHD) flows are critical to understanding complex flow phenomena in ducts of liquid metal blankets, in particular those that utilize eutectic alloy lead–lithium as breeder/coolant, such as self-cooled, dual-coolant and helium-cooled lead–lithium blanket concepts. The primary goal of MHD experiments at UCLA using the liquid metal flow facility called MaPLE (Magnetohydrodynamic PbLi Experiment) is to address important MHD effects, heat transfer and flow materials interactions in blanket-relevant conditions. The paper overviews the one-of-a-kind MaPLE loop at UCLA and presents recent experimental activities, including the development and testing of high-temperature PbLi flow diagnostics and experiments that have been performed since the first loop operation in 2011. We also discuss MaPLE upgrades, which need to be done to substantially expand the experimental capabilities towards a new class of MHD flow phenomena that includes buoyancy effects.

Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO.This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of the equatorial outboard module has been analyzed with CAST3M in case of accident, according to RCC-MRx, resulting in relatively good behavior of the box. Additionally analyses have been performed in order to check the good behavior of the tie rods modules’ attachments system in normal condition showing promising results.

Supercritical CO2 Brayton power cycles for DEMO fusion reactor based on Helium Cooled Lithium Lead blanket

Fusion energy is one of the most promising solutions to the world energy supply. This paper presents an exploratory analysis of the suitability of supercritical CO2 Brayton power cycles (S-CO2) for low-temperature divertor fusion reactors cooled by helium (as defined by EFDA). Integration of three thermal sources (i.e., blanket, divertor and vacuum vessel) has been studied through proposing and analyzing a number of alternative layouts, achieving an improvement on power production higher than 5% over the baseline case, which entails to a gross efficiency (before self-consumptions) higher than 42%. In spite of this achievement, the assessment of power consumption for the circulating heat transfer fluids results in a penalty of 20% in the electricity production. Once the most suitable layout has been selected an optimization process has been conducted to adjust the key parameters to balance performance and size, achieving an electrical efficiency (electricity without taking into account auxiliary consumptions due to operation of the fusion reactor) higher than 33% and a reduction in overall size of heat exchangers of 1/3. Some relevant conclusions can be drawn from the present work: the potential of S-CO2 cycles as suitable converters of thermal energy to power in fusion reactors; the significance of a suitable integration of thermal sources to maximize power output; the high penalty of pumping power; and the convenience of identifying the key components of the layout as a way to optimize the whole cycle performance.

Technologies and modelling issues for tritium processing in the European Test Blanket Systems and perspectives for DEMO

One of the main objectives of the experimental campaign on the Test Blanket Systems (TBS) in ITER is the demonstration of the efficient processing of the tritium generated in the Test Blanket Module (TBM).On the other side, efficient tritium processing in a TBS has deep implications on: (i) safe operation of TBS itself and whole ITER system; (ii) successful development and validation of tritium transport modelling codes; (iii) demonstration of DEMO relevancy of tritium processing technologies.This work describes various aspects of HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS activities related to TBS tritium management. After a short description of HCLL and HCPB blanket concepts and related TBS, the paper contains:1.a presentation of the key tritium processing technologies in the current design baseline of the European TBS;2.a discussion on the DEMO relevancy of some specific TBS tritium processing technologies;3.an overview on the activities related to the tritium transport modelling tools that will be validated along the development of the TBM project, including experimental campaign in ITER, and used for supporting the DEMO Breeding Blanket design.These three items are connected each other since tritium-related data, generated through the experimental campaign in ITER and interpreted through suitable modelling tools, will be one of the most significant outcomes in support of the breeding blanket design for DEMO and beyond.

Development of the Helium Cooled Lithium Lead blanket for DEMO

The Helium Cooled Lithium Lead (HCLL) blanket is one of the candidate European blanket concepts selected for the DEMOnstration fusion power plant that should follow ITER. In a fusion power plant, the blanket is one of the key components because of its impact on the plant performance, availability, safety and economics. In 2012, the European Fusion Development Agreement (EFDA) agency issued new specifications for DEMO: this paper describes the work performed to adapt the previous 2007 HCLL-DEMO blanket design to those specifications.A new segmentation has been defined assuming straight surfaces for all blanket modules. Following the Multi Module Segment (MMS) option, all modules are attached to a common back supporting structure which also serves as manifold for Helium and PbLi distribution. A detailed CAD design of the central outboard module has been defined. Thermo-hydraulic and thermo-mechanical analyses on of the First Wall and Breeder Zone have been carried out. For the attachment of the modules to the common backplate, a new solution based on the use of Tie Rods, derived from the design of the corresponding HCLL Test Blanket Module for ITER, has been proposed. This paper also identifies the priorities for further development of the HCLL blanket design.

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).