LiPb Blanket Research Articles

Design status and development strategy of China liquid lithium–lead blankets and related material technology

A series of fusion reactors (named FDS series) have been designed and assessed in China, with four types of liquid lithium lead blankets including the RAFM steel-structured He-cooled quasi-static LiPb tritium breeder (SLL) blanket, the RAFM steel-structured He–LiPb dual-cooled (DLL) blanket, the RAFM steel-structured refractory material thermally-insulated high temperature LiPb (HTL) hydrogen production blanket and the RAFM steel or optionally the austenitic stainless steel-structured He–LiPb dual-cooled high level waste transmutation (DWT) blanket. To demonstrate and validate the feasibility of the candidate blankets for fusion energy application, the three-phases-strategy of TBM (test blanket module) development, i.e. material R&D and out-of-pile experimental mockup, EAST-TBM and ITER-TBM have been proposed. A brief overview of the four types of LiPb blanket designs and their goals are given. Material technology requirement and development strategy are also presented in this paper.

Conceptual design activities of FDS series fusion power plants in China

A series of fusion power plants (named FDS series) have been designed and assessed for the examination of the feasibility and the safety, environmental and economical potential of fusion with emphasizing the blanket design optimization on neutronics, thermal-hydraulics, electro-magnetics, material, structural performance analyses in China. Four concepts have been developing, which are the fusion-driven subcritical system (FDS-I) with the goal of transmutation of the long-lived nuclear wastes and breeding of fissile nuclear fuels, the fusion power reactor (FDS-II) with the goal of electricity generation, the fusion-based hydrogen production reactor (FDS-III) and the spherical tokamak-based compact reactor (FDS-ST). Four types of blankets, including the He/LiPb dual-cooled high level waste transmutation (DWT) blanket, the He single-cooled LiPb tritium breeder (SLL) blanket and the He/LiPb dual-cooled (DLL) blanket, the high temperature liquid LiPb (HTL) blanket, and Test Blanket Modules (TBMs) and the testing strategy have been studied. R&D status on materials, design and analysis tools are also presented.

Design of auxiliary system for the dual functional lithium–lead test blanket module in ITER

LiPb auxiliary system is an important external loop equipment of test blanket module (TBM) for liquid LiPb blanket design. According to the requirement of testing strategy of DFLL (dual functional lithium lead) TBM, which could demonstrate the technology of the SLL (single-coolant lithium lead) and the DLL (dual-coolant lithium lead) blankets, a conceptual design of the LiPb ancillary system is done, which could serve to achieve the dual functions. The auxiliary system can be used to achieve the recovery of tritium, heat rejection, monitoring of impurity in LiPb, LiPb purification and so on. It consists of detritiation unit, LiPb to He heat exchanger, plugging indicator, LiPb purification system and pumping system, etc.

Neutron imaging of fiber-reinforced materials

Glass-fiber-reinforced plastic laminates used for the insulation of Toroidal Field (TF) magnet-coils and fiber-reinforced silicon carbide ceramic composites used as structural material for the self-cooled Pb-17Li blanket module are attractive candidate materials for fusion reactors because of their high performance under extreme conditions. Porosity, which depends on the manufacturing process, and swelling of fiber-reinforced materials due to the high flux of radiation are the main problems. The aim of this study is to describe the experimental procedures of different imaging methods, and also to decide the most efficient imaging method for the investigations of the complex microstructure of fiber-reinforced materials. In this work, the fiber-reinforced composites were inspected with neutron and X-ray radiographies at ATI-Vienna and also at PSI-Villigen. A contrast enhancement at the edges can be achieved by means of phase contrast neutron radiography (NR), which is based on the wave properties of neutrons and arises from the neutron refraction (rather than attenuation). Elements having different refractive index within a sample cause a phase shift between coherent neutron waves. The degree of coherence can be determined by means of the coherence pattern caused by the sample, when a point source (pinhole) is used and the distance between source and sample is varied.

Component lifetime comparison and waste volume in CLiFF Sn/Flibe and Sn/LiPb blankets

The thin Convective Liquid Flow First Wall (CLiFF) concept is one of the liquid FW concepts investigated in the Advanced Power Extraction (APEX) study for high power density application. Liquid tin has been suggested as the 2 cm thick front flowing liquid layer because of its low vapor pressure. Two choices were selected for the conventional blanket that follows the thin liquid wall (LW), namely: (1) LiPb/SiC blanket, and (2) Flibe/SiC blanket. Lithium is enriched to 90% Li-6 in the first blanket option and to 25% Li-6 in the second option (with 10 cm-thick beryllium front zone). Due to the superior attenuation characteristics of Flibe over LiPb, this impacted the lifetime of the SiC structure used in both options. In this paper, we assessed the lifetime of the SiC structure in the FW/Blanket and the shield in both blanket options. The end-of-life limit of 200 dpa is assumed (corresponding to ∼3% burn-up). The frequency of replacement of each component is estimated based on 30-years plant lifetime. Comparison is made for the waste volume of replaced components in each option. It is shown that the shield can last the plant lifetime in the Flibe blanket while part of the shield in the LiPb blanket will require replacement. The frequency of replacing the FW/Blanket with the LiPb blanket option is twice as much as in the Flibe blanket option. This is translated to a total volume of disposed structure at plant end-of-life from the entire FW/B/shield system that is larger by ∼60%.

Progress on the TAURO blanket system

TAURO is a self-cooled Pb17Li blanket for Fusion Power Reactors (FPRs) using SiC f/SiC composites structures. It has been developed with the objective to achieve both passive safety and high thermal efficiency. The Pb17Li outlet temperature exceeds 850 °C. A Brayton cycle is envisaged for power conversion and the possibility of hydrogen production is addressed. This paper recalls the main features of the TAURO blanket, and presents the performed investigations on a range of potential operations conditions with regard to plant thermal efficiency and H 2-production. It also presents the recent advances made in the design methodologies of composite structures for fusion components involving the implementation in finite-element codes of material models capable of describing the complex non-linear mechanical behavior of these composites and the definition of new resistance criteria. It is possible with this new model to better treat the thermo-mechanical results. Indeed, stresses in each direction are compared directly with allowable limits for tension, compression and shear stresses in each orthotropic direction. Such a model has been implemented in the FEM code CASTEM 2000 developed at CEA. Fabrication aspects are discussed based on a brazing process currently under development and briefly presented. Moreover, the advantages of using newly-produced advanced fibers in SiC f/SiC composites in relation with the main issues (namely irradiation stability and thermal conductivity) are assessed. The potential of a liquid metal cooled divertor using SiC f/SiC as structural material, in line with the TAURO reactor characteristics is also investigated. A circuit in series could be envisaged, where the Pb17Li flows at first in the divertor and then it passes in the blanket, which allows increasing the blanket inlet temperature. Calculations on a castellated square tube basic concept showed a reasonable maximum allowable heat flux of about 5 MW/m 2.

Further improvements of the water-cooled Pb–17Li blanket

The water-cooled lithium–lead (WCLL) blanket is based on reduced-activation ferritic–martensitic steel as the structural material, the liquid alloy Pb–17Li as breeder and neutron multiplier, and water at typical PWR conditions as coolant. It was developed for DEMO specifications and shall be tested in ITER. In 1999, a reactor parameter optimization was performed in the EU which yielded improved specifications of what could be an attractive fusion power plant. Compared to DEMO, such a power reactor would be different in lay-out, size and performance, thus requiring to better exploit the potential of the WCLL blanket concept in conjunction with a water-cooled divertor. Several new approaches are currently under evaluation. This paper outlines several specific modifications, it highlights progress made on various issues and outlines the R&D work which is still required to define an improved reference design for the WCLL concept.

Potential and limits of water-cooled Pb–17Li blankets and divertors for a fusion power plant

Blankets and divertors are key components of a fusion power plant. They have a large impact on the overall plant design, its performance and availability, and on the cost of electricity. The water-cooled Pb–17Li (WCLL) blanket uses reduced activation ferritic–martensitic steel as structural material. It was previously validated under numerous aspects such as TBR, mechanical and thermo-mechanical stability, thermal–hydraulics, MHD, safety and others. This was done assuming the specifications for a European DEMOnstration reactor which were fixed back in 1989. A WCLL blanket would best be combined with a water-cooled divertor so that a single coolant could be used for the entire reactor. Several divertor designs were proposed recently. This paper investigates the applicability of the WCLL blanket concept and a water-cooled divertor in attractive power reactors with increased power densities compared with DEMO.

A helium-cooled blanket design of the low aspect ratio reactor

An aggressive low aspect ratio scoping fusion reactor design (C.P.C. Wong, R. Cerbone, E.T. Cheng, R.L. Miller, R.D. Stambaugh, Proc. of 17th IEEE/NPSS Symp. on Fusion Engineering, pp. 1053) indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of >4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 430°C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m 2, respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650°C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

Hydrogen isotope permeation through and inventory in the first wall of the water cooled Pb–17Li blanket for DEMO

The hydrogen isotope transport through the first wall between plasma and first wall coolant has been investigated. The time dependence of hydrogen isotope permeation through the first wall from plasma into water coolant has been calculated. The hydrogen isotope inventory (both mobile and trapped hydrogen isotopes) in the first wall has been evaluated. The influence of temperature gradient, surface conditions, isotopic effect and trapping in the ion- and neutron-induced defects on the hydrogen isotope permeation and inventory has been considered.

Design of a tritium breeding blanket for volumetric neutron source

A water-cooled and austenitic stainless-steel-structured breeding blanket system is designed for a volumetric neutron source (VNS), based on a steady-state tokamak device. The designed VNS with super-conducting coils has a 4.5 m main radius and a total of 300 MW in fusion power. It yields the tritium consumption of approximately 10 kg per year with 50% availability. It is unrealistic to assume that this amount of tritium could be supplied by other nuclear facilities, therefore, tritium breeding with the VNS device itself should be accommodated. The capability of tritium breeding for the Li-Pb breeder is examined. Compared to ceramic breeding materials, in general, the Li 17Pb 83 breeder has the potential for a higher tritium breeding ratio. In addition, the compatibility of Li-Pb with water coolant and with structure material at low temperature and low flow velocity may be acceptable. The calculated local TBR in the present design reached 1.4 or more in the Li-Pb breeder with a sufficient thickness of the breeding zone. The temperature distribution in the breeding region of the Li-Pb blanket was controlled to avoid corrosion of the structure material, using a double-walled cooling panel. The net TBR might attain the acceptable value when only the outboard Li-Pb blanket is used, since the contribution of outboard blanket to tritium breeding is higher than its coverage.

Design development and manufacturing sequence of the European water-cooled Pb-17Li test blanket module

In 1996, the European Community started the development of a water-cooled Pb-17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the Basic Performance Phase prior to d–t operation. The test module is designed to be representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined.

Effects of heat treatment and cooling rate on microstructural and mechanical characteristics of aluminide coatings on MANET steel

R & D studies concerning Pb-17Li blankets for DEMO suggest the use of coatings either as tritium permeation barriers for the water cooled blanket or as electrical insulators for the self cooled blanket. The production of coatings resistant to the thermal and mechanical stresses typical of the blanket operating conditions is the target. The present work describes the results of an experimental programme aimed at evaluating the effects of heat treatment and cooling rate on the microstructural and mechanical characteristics of aluminide coatings deposited on MANET steel (DIN 1.4914) by the hot-dipping process. The temperature of the post-deposition heat treatment that allows obtaining structures like FeAl and α-Fe(Al), instead of the more brittle FexAly intermetallics, has been identified. The coatings have been characterized by optical metallography, microhardness measurements, scanning electron microscopy (SEM), energy dispersive X-ray analysis (EDS) and X-ray diffraction on the surface.

Tritium Confinement Requirements for the Water-Cooled Pb-17Li Blanket for DEMO

In the water-cooled liquid Pb-17Li blanket concept for DEMO the limitation of tritium permeation from the breeder material into the cooling water will be required. In order to find out on what conditions this tritium permeation remains within reasonable limits, a 1-d FEM code was developed which evaluates the tritium partial pressure in Pb-17Li, the tritium inventory in the blanket material, and the tritium permeation from the Pb-17Li into the cooling water as a function of the permeation reduction factor of a barrier and the efficiency of the tritium extraction from Pb-17Li. With a parametric study the conditions were identified which allow a permeation rate of as little as 1 g.d -1 without pushing the requirements for permeation barriers and extraction efficiencies excessively far. An example is a barrier with a permeation reduction factor of 75 together with an extractor efficiency of approximately 83%. In these conditions the expected tritium inventory is attributed to approximately one third to the martensitic blanket structure (some ten grams) while two thirds will be found in the Pb-17Li. These inventory values are two orders of magnitude lower than in solid breeder blankets and are thus not considered a critical issue.

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

Magnetohydrodynamic investigations of a self-cooled Pb-17Li blanket with poloidal-radial-toroidal ducts

Abstract For self-cooled liquid metal blankets, the magnetohydrodynamic (MHD) pressure drop and velocity distributions are considered as critical issues. This paper summarizes MHD work performed for a DEMO-related Pb-17Li blanket, where the coolant flows downwards in rear poloidal ducts; turns around by 180° at the blanket bottom; is diverted from poloidal ducts into short radial channels which feed to toroidal First wall coolant ducts; flows through the subsequent radial channels; is collected again in poloidal channels and leaves the blanket segment at the blanket top. To reduce the pressure drop and to decouple electrically parallel channels, flow channel inserts are used for all the ducts except the first wall ducts. A previous pressure drop assessment resulted in significant values for duct geometries with flow distribution or collection, and multichannel effects for the system of U-bends. As a result of the uncertainty of these assessments, corresponding investigations were carried out recently. Characteristic results are presented in this paper. It is shown that, for both geometries, the pressure drops are considerably lower than those previously assessed. First results from experiments on the velocity distribution in a radial-toroidal-radial U-bend are also presented. Here, it is shown that, with an increasing interaction parameter, the liquid preferentially flows close to the First wall. Additionally, a pair of strong vortices was observed in a toroidal duct. Both effects are supposed very favourable for heat transfer.

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

Abstract The question of the tritium recovery in water-cooled Pb 17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb 17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pbl7Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit.

Tritium/hydrogen barrier development

A review of the hydrogen permeation barriers which can be applied to the structural metals used in fusion power plants is presented. Both implanted and chemically available hydrogen isotopes must be controlled in fusion plants. The need for permeation barriers appears strongest in Pb-17Li blanket designs, although barriers are also necessary for other blanket and coolant systems. Barriers which provide greater than a 1000- fold reduction in the permeation of structural metals are desired. In laboratory experiments, aluminide and titanium ceramic coatings provide permeation reduction factors (PRFs) of 1000 to over 100 000 with a wide range of scatter. The rate-controlling mechanism for hydrogen permeation through these barriers may be related to the number and type of defects in the barriers. Although these barriers appear robust and resistant to liquid metal corrosion, irradiation tests which simulate blanket environments result in very low PRFs in comparison with laboratory experiments, i.e. less than 150. It is anticipated from fundamental research activities that the radiation- and electric-field-induced enhancement of hydrogen diffusion in oxides may contribute to the lower PRFs during in-reactor experiments.

European Reference Design of the Water-Cooled Lithium-Lead Blanket for a Demonstration Reactor

The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European Union for DEMO-relevant design and R&D activities. This paper gives a presentation of the reference...

Influence of a magnetic field on the corrosion of austenitic and martensitic steels by semi-stagnant Pb17Li

The influence of a magnetic field on the compatibility of 316L austenitic and 1.4914 martensitic steels with Pb17Li has been studied in conditions simulating the special features of the water-cooled Pb17Li blanket (low Pb17Li velocity, significant radial thermal gradient and short distances between hot and cold zones). In the 420–475°C temperature range, the results show an increase of the corrosion rate in the presence of a magnetic field. This increase is about 50% for 316L steel and 30% for 1.4914 martensitic steel. Moreover the magnetic field induces a loss of symmetry in the deposition process: the amount of recovered deposit is greater in the direction parallel to the magnetic field than in the perpendicular one.