Water Cooled Lithium Lead Blanket Research Articles

Thermo-structural design of the European DEMO water-cooled blanket with a multiscale-multiphysics framework

The Water-Cooled Lithium-Lead (WCLL) blanket concept is one of four possible embodiments under consideration for the European Demonstration Power Plant DEMO, beyond ITER. It is based on a water-cooled first wall and blanket structure with the RAFMS (Reduced Activation Ferritic/Martensitic Steel, Eurofer 97) as structural material and PbLi as neutron multiplier and tritium breeder. We first develop a multiphysics modeling framework in order to optimize the design and achieve long blanket lifetime, maintainability, and high reliability. The approach is demonstrated by coupling Computational Fluid Dynamics (CFD), heat transfer and solid mechanics. The multiphysics framework is then utilized to provide the loading conditions for a multiscale design approach particularly focused on modeling the effects of both steady-state and operational transients on the structural integrity of the First Wall (FW). The methodology builds on the ITER Structural Design Criteria for In-Vessel Components (ISDC-IC), then extending the criteria to progressively incorporate continuum plasticity models and microstructure-based representation of deformation and fracture. We finally introduce advanced fracture mechanics concepts based on a Materials-Specific Failure Assessment Diagram (MS-FAD), in which precise calculations of the J-integral of an elasto-plastic material is required. This is especially important for the water-cooled design because of the limited fracture toughness of Eurofer 97 below 300 °C. The methodology has been applied to screen the FW/Blanket performance of a range of steady state heat flux assumed to be applied during the plasma pulse flat top, and up to 0.8 MW/m2. It has become clear that the radiation steady state load will be lower than the values screened here, while the local heat flux due to particle loads will lead to global higher values than the values investigated here. The methodology will be further fine-tuned to specific loading conditions in order to assess actual performance limits of the DEMO WCLL BB.

Neutronic analyses in support of the WCLL DEMO design development

In the frame of the EUROfusion Consortium programme, the Water Cooled Lithium Lead (WCLL) option has been chosen as a candidate for the breeding blanket (BB) of the European fusion power demonstration plant (DEMO) conceptual design. Neutronic analyses play a fundamental role in the development of the WCLL blanket, providing guidelines for its design based on the evaluation of the nuclear performances.A detailed three-dimensional MCNP model of the latest WCLL layout has been generated and integrated in a DEMO MCNP generic model suitably designed for neutronic analyses. Three-dimensional neutron and gamma transport simulations have been performed using the MCNP5v1.6 Monte Carlo code and JEFF 3.2 nuclear data libraries, in order to assess the WCLL-DEMO tritium self-sufficiency and the shielding capabilities of the breeding blanket/manifold system to protect the vacuum vessel and toroidal field coils. Furthermore, radial profiles of the neutron flux, nuclear heating, neutron damage and he-production have been assessed in the inboard and outboard equatorial planes.The outcome of the present study highlights the potential and suitability of the WCLL breeding blanket for the application to DEMO, both in terms of tritium production and shielding performances.

Preliminary System Modeling for the EUROfusion Water Cooled Lithium Lead Blanket

The Water Cooled Lithium Lead (WCLL) blanket is one of the four breeder blanket technologies under consideration within the framework of the EUROfusion Consortium activities. The aim of this work is to develop a preliminary model that can track tritium concentration and tritium fluxes along each part of the WCLL blanket and its ancillary systems at any time.Because of tritium’s nature, the phenomena of diffusion, dissociation, recombination and solubilization have been taken into account when describing the tritium behavior inside the lead-lithium channels, the structural materials and the water coolant circuits. The simulations have been performed using the object oriented modeling software EcosimPro.Results have been obtained for the pulsed generation scenario of the European demonstration power plant (DEMO). The tritium inventory in every part of the blanket has been computed. Permeation rates have been calculated as well allowing to know how much tritium ends up in the coolant system and how much remains in the liquid metal. The amount of tritium extracted from the lead-lithium loop has been also obtained. All this information allows having a global perspective of tritium behavior all over the blanket at any time.The model provides valuable information for the design of the WCLL blanket. More complex upgrades are planned to be implemented based on this model in future stages of the EUROfusion project.

Dynamic thermal-hydraulic modelling of the EU DEMO WCLL breeding blanket cooling loops

A global thermal-hydraulic model of the EU DEMO tokamak, based on an object-oriented approach using the Modelica language, is currently under development at Politecnico di Torino. The first module of the global model will simulate the dynamics of the blanket cooling system and it will be able to investigate different coolant options and different cooling schemes, adapted to the different blanket systems currently under development in the Breeding Blanket project. Following the development of the module related to the Helium-Cooled Pebble Bed blanket, this paper presents the dynamic model of the Water-Cooled Lithium Lead (WCLL) blanket, designed at ENEA Brasimone, together with the first results obtained in steady-state conditions.

Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-cooled lithium-lead has the highest total decay heat at longer decay times in comparison to the helium-cooled design which has the lowest total decay heat. In addition, major nuclides were identified for water-cooled lithium-lead in W armour, Eurofer, and LiPb. In addition, great attention has been dedicated to the analysis of the decay heat and activity both from the different water-cooled lithium-lead blanket modules for the entire reactor and from each water-cooled lithium-lead blanket module separately. The neutron induced activation and decay heat at shutdown were calculated by the FISPACT code, using the neutron flux densities and spectra that were provided by the preceding MCNP neutron transport calculations.

Simulation study of pressure trends in the case of loss of coolant accident in Water Cooled Lithium Lead blanket module

The water–lithium lead interaction implies a direct energy release, which leads to temperature and pressure increase, due to a combined thermal and chemical reaction, and an indirect form of energy release, the hydrogen production, due to secondary chemical reaction involving the initial reaction products. Review and understanding of the knowledge acquired in past studies, experimental works and numerical activities are needed in view of the renewed interest in the Water Cooled Lithium Lead blanket concept and safety issues connected with the fusion reactor design. This paper presents a review of the studies carried out in the past to characterize the potential safety concerns associated with the use of water and lithium-lead eutectic alloy, the main experimental campaigns, and numerical simulations of BLAST Test No. 5 performed by SIMMER-III code. As results, no code was found able to perform a satisfactory post-test analysis of separate effect experiments, without engineering assumptions. Therefore, a code model for the exothermic reaction and hydrogen production, and experimental data are needed for solving the WCLL blanket safety issues associated with the water–PbLi interaction.

Analysis of the thermo-mechanical behaviour of the DEMO Water-Cooled Lithium Lead breeding blanket module under normal operation steady state conditions

Within the framework of DEMO R&D activities, a research cooperation has been launched between ENEA, the University of Palermo and CEA to investigate the thermo-mechanical behaviour of the outboard equatorial module of the DEMO1 Water-Cooled Lithium Lead (WCLL) blanket under normal operation steady state scenario. The research campaign has been carried out following a theoretical–computational approach based on the Finite Element Method (FEM) and adopting a qualified commercial FEM code. In particular, two different 3D FEM models (Model 1 and Model 2), reproducing respectively the central and the lateral poloidal–radial slices of the WCLL blanket module, have been set up. A particular attention has been paid to the modelling of water flow domain, within both the segment box channels and the breeder zone tubes, to simulate realistically the coolant-box thermal coupling. Results obtained are herewith reported and critically discussed.

Preliminary RAMI analysis of WCLL blanket and breeder systems

DEMO will be a prototype fusion reactor designed to prove the capability to produce electrical power in a commercially acceptable way. One of the key factors in that endeavor is the achievement of certain level of plant availability. Therefore, RAMI (Reliability, Availability, Maintainability and Inspectability) will be a key element in the engineering development of DEMO.Some studies have been started so as to develop the tools and models to assess different design alternatives from RAMI point of view. The main objective of these studies is to be able to evaluate the influence of different parameters on DEMO availability and to focus the critical parts that should be further researched and improved in order to develop a high-availability oriented DEMO design.A preliminary RAMI analysis of the Water Cooled Lithium-Lead (WCLL) blanket and breeder concept for DEMO has been developed. The amounts of single elements that may fail (e.g. more than 180,000 C-shaped tubes) and the mean down time associated to failures inside the vacuum vessel (around 3 months) have been highlighted as the critical parameters influencing the system availability. On the other hand, the necessary developments of tools/models to better represent the system performance have been identified and proposed for future work.

Optimization of the first wall for the DEMO water cooled lithium lead blanket

The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity.The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0MWm−2 with respect to the Eurofer temperature limit.An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet temperatures equal to 285°C/325°C respectively and based on geometric design parameters such as channel pitch, diameter of pipes and thicknesses. It became clear through the optimization that the advantages of a waved FW are diminished. Given the manufacturing issues of that concept, the waved FW was therefore not pursued further. Even if the optimization study shows that the maximum heat load could in principle be as high as 2.53MWm−2, it is reduced to 1.57MWm−2 when additional constraints are introduced in order not to affect corrosion, manufacturability and Tritium Breeding Ratio in normal condition such as a coolant velocity ≤8m/s, pipe diameter ≥5mm and a total FW thickness ≤22mm.However it is important to note that the FW channels currently fulfill additional functions and are therefore not optimized “at all cost” regarding heat load capacity and the paper points out some recommendations against missing assumptions.

Assessment of the Thermo-mechanical Performances of a DEMO Water-Cooled Liquid Metal Blanket Module

Within the framework of DEMO R&D activities, a research cooperation has been launched between ENEA-Brasimone, CEA-Saclay and the University of Palermo to investigate the thermo-mechanical behaviour of the outboard equatorial module of the DEMO1 Water-Cooled Lithium Lead (WCLL) blanket, both under normal operation and over-pressurization steady state scenarios. The research campaign has been carried out following a theoretical-computational approach based on the finite element method (FEM) and adopting a qualified commercial FEM code. In particular, two different three-dimensional FEM models of the WCLL blanket module have been set-up to be used for normal operation and over-pressurization analysis, respectively. The former reproduces realistically the WCLL blanket module central poloidal–radial region, including one breeder cell in the toroidal direction and all the five cells in the radial one. The latter simulates the entire module segment box, without including the breeder and its cooling tubes. Two set of uncoupled steady state thermo-mechanical analyses have been carried out with reference to the investigated loading scenarios. In particular, under normal operation scenario (level A) the module has been supposed to undergo both 15.5 MPa coolant pressure on its cooling channels walls and thermal deformations due to the flat-top plasma operational state thermal field, while under over-pressurization scenario (level D) it has been assumed to experience 15.5 MPa coolant pressure on its segment box internal walls while operating at normal operation thermal level. Results obtained are presented and critically discussed according to the SDC IC code.

Development of the water cooled lithium lead blanket for DEMO

The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant.In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications.Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant.A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood.The priorities for further development of the WCLL blanket concept are identified in the paper.

Pb-16Li/water interaction: Experimental results and preliminary modelling activities

The Water Cooled Lithium Lead (WCLL) blanket is based on the eutectic liquid alloy Pb16Li as breeder material and neutron multiplier, and pressurised water as coolant. The liquid breeder flows at few mm/s in the blanket module while the pressurised water is circulated inside double-wall tubes.In spite of the adoption of double-wall tubes for the coolant, the probability of a water large leak because of a tube rupture accident cannot be considered negligible. As a consequence, the Pb16Li/water interaction due to a large break in one or more cooling tubes still remains one of the biggest concerns for this blanket concept.This paper reports the results of three experimental tests on Pb16Li/water interaction carried out at ENEA-Brasimone operating the LIFUS 5 facility. Water was injected into the reaction tank, containing Pb16Li at 330°C, at a pressure of 155bar with different values of sub-cooling and with different free volumes in the reaction system. In addition, post test analyses with SIMMER III code are presented in order to compare the pressure evolution measured during the experiments with that calculated by the code.

Blanket Concept of Water-Cooled Lithium Lead with Beryllium for the SlimCS Fusion DEMO Reactor

As an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. In SlimCS, the net tritium breeding ratio (TBR) supplied from WCLL blanket was not enough because the thickness of blanket in SlimCS was limited to about 0.5 m so as to allocate the conducting shell position near the plasma for high beta access and vertical stability of plasma. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier to reach a required TBR (≥ 1.05). Considering the operating temperature of blanket materials, a double pipe structure was adopted. The nuclear and thermal analysis were carried out by a nuclear-thermal-coupled code, ANIHEAT and DOHEAT so that blanket materials were appropriately arranged to satisfy the acceptable operation temperatures. The temperatures of materials were kept in appropriate range for the neutron wall load Pn = 5 MW/m2. It was found that the local TBR of WCLL with Be blanket was comparable with that of solid breeder blanket.

Study on thermodynamic characteristic and optimization of steam cycle system in IGCC : Jiang, L. et al. Energy Conversion and Management, 2002, 43, (9–12), 1339–1348

In this paper, a water-cooled divertor concept is assessed and optimized according to operating conditions foreseen for the water-cooled lithium lead (WCLL) blanket concept in the frame of the European Power Plant Conceptual Study (PPCS). In a power reactor, the potential performances of a divertor concept are judged not only by its capability to sustain high heat flux, but also by the availability of the extracted heat for power conversion. A water-cooled divertor is the preferred option when used with a WCLL blanket because it simplifies the reactor coolant system design and reduces the required R&D by using the same coolant for all components. A Cu-alloy tube is used as heat sink, armored by a tungsten monoblock. A swirl tape is inserted in the tube, as a turbulence promoter. This concept needs limited extrapolations from today's plasma physics assumptions insofar it allows high surface heat fluxes (up to 15 MW m−2). At the same time, it permits the divertor power to contribute in the thermodynamic cycle of the plant with an overall maximum gross thermal efficiency of about 33% similar to that of a pressurized water reactor.

In-pile performance of a double-walled tube and a tritium permeation barrier

In two recent irradiation experiments in the HFR Petten, tritium permeation rates through representative materials to be used as cooling tubes of the water-cooled lithium-lead blanket have been measured in-pile. These latest experiments in the EXOTIC 8 series (E 8.9 and E 8.10) are made of a double wall tube (DWT) and a T91 tube with an Fe–Al/Al 2O 3 layer acting as tritium permeation barrier (TPB). These tubes contain annular pebble beds of ceramic breeder materials for the helium-cooled pebble bed concept blanket as tritium breeding material. Both experiments are built up of two concentric and independently purged containments allowing on-line tritium release rate and permeation rate measurements. In-pile operation has ended in March 2001 after 450 full power days and resulted in an irradiation damage of approximately 2.6 and 3.2 dpa, respectively in T91 steel. This paper reports on the experimental results obtained for in-pile tritium permeation and discusses the influence of purge gas compositions, temperature and irradiation on tritium permeation through the DWT and TPB.

Optimization of a water-cooled divertor for the European power plant conceptual study

In this paper, a water-cooled divertor concept is assessed and optimized according to operating conditions foreseen for the water-cooled lithium lead (WCLL) blanket concept in the frame of the European Power Plant Conceptual Study (PPCS). In a power reactor, the potential performances of a divertor concept are judged not only by its capability to sustain high heat flux, but also by the availability of the extracted heat for power conversion. A water-cooled divertor is the preferred option when used with a WCLL blanket because it simplifies the reactor coolant system design and reduces the required R&D by using the same coolant for all components. A Cu-alloy tube is used as heat sink, armored by a tungsten monoblock. A swirl tape is inserted in the tube, as a turbulence promoter. This concept needs limited extrapolations from today's plasma physics assumptions insofar it allows high surface heat fluxes (up to 15 MW m −2). At the same time, it permits the divertor power to contribute in the thermodynamic cycle of the plant with an overall maximum gross thermal efficiency of about 33% similar to that of a pressurized water reactor.

A water cooled lithium–lead blanket without tritium permeation barriers: feasibility and economical analysis

In the reference design of the water cooled lithium–lead (WCLL) blanket for DEMO, tritium permeation into the primary coolant must be lowered as much as possible to keep the coolant specific activity not higher than 1 Ci/kg. This requires tritium permeation barriers to be applied on the cooling tubes in the blanket. Taking into account the aggressive environments and the huge surface to be coated, the technology of permeation barriers appears as a tough challenge. To evaluate if it is feasible to lower the requirements on the performance of tritium permeation barriers, thus accepting a higher tritium permeation rate into the coolant and imposing a higher capacity in extracting tritium from the primary coolant, a conceptual study has been performed, mainly focusing on economical aspects. The results, which seem up to now encouraging, are presented and discussed.

Hydrogen permeation through tritium permeation barrier in Pb–17Li

One of the main problems in the development of water cooled lithium lead (WCLL) DEMO fusion reactor is the reduction of the tritium permeation from the Pb–17Li, or the plasma, into the cooling water. The control of tritium losses is an important issue in fusion technology because of its safety and operational implications. This goal can be achieved using a tritium permeation barrier (TPB). The use of aluminium rich coatings, which forms Al 2O 3 at their surface, has been selected as reference solution for WCLL blanket in order to produce reliable TPB. The hot dipping process is one of the two candidates for the production of coatings on large blanket segments. The effectiveness of hot dipped aluminium coating on MANET II steel has been verified in gas phase and in liquid Pb–17Li, using a test apparatus named Corelli II. The permeation rate measured in gas phase is one order of magnitude lower than the one in liquid metal phase. Performing SEM-EDS analysis on the specimen, it was observed that micro cracks on the coating surface were present. The permeation curves in gas and Pb–17Li are reported and discussed. The characteristics of the new experimental device Vivaldi will be briefly described.

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.

General Ophthalmologic Examination

Caring for patients with eye emergencies will always challenge even the most experienced emergency clinician. The threat of loss of sight is anxiety provoking for patients and requires the physician to act in an organized and expeditious manner. By avoiding these pitfalls and examining each patient in a systematic and comprehensive fashion, the physician can ensure the best possible outcome. The ensuing articles describe the approach to diagnosing and treating significant eye emergencies in greater detail.