Cooled Lithium Lead Breeding Blanket Research Articles

RELAP5/Mod3.3 thermal-hydraulics characterization of the steam generator mock-up during operational transients in STEAM facility in support of the design of the DEMO WCLL BoP

The Water Cooled Lithium Lead Breeding Blanket (WCLL BB) is a key candidate for the driver blanket of the European DEMO reactor, progressing toward its Conceptual phase by the end of 2027. To assess different water and lithium-lead technologies for the WCLL BB and Balance of Plant (BoP) systems, the Water-thermal-HYDRAulic (W-HYDRA) experimental platform is under development at the ENEA Brasimone Research Centre. Among the facilities constituting the new W-HYDRA multipurpose infrastructure, STEAM is going to experimentally investigate the DEMO WCLL BoP thermal-hydraulics, focusing on the Steam Generator (SG) of the Primary Heat Transfer Systems (PHTS), to qualify its performances and suitability under its unconventional operation. The paper aims at supporting the thermal-hydraulic characterization of the Steam Generator mock-up during the sudden power variations typical of a pulsed fusion reactor. The analyzed selected scenario is the operational transient dwell-pulse-dwell, which determines high thermal cycling and correspondent high thermo-mechanical stresses on the primary side components. Two control logics with their relative drawbacks have been analyzed with a RELAP5/Mod3.3 1-D model, the first regulating the primary side average temperature, the second monitoring the minimum one. The comparison of the two systems highlighted that neither approach leads to hazardous conditions for the facility. However, while the average temperature controller is characterized by reduced thermal stresses on the components, the minimum temperature controller is characterized by higher thermal gradients for the primary loop. Both methodologies will be tested in the dedicated experimental campaign, aiming at yielding insights and evaluations concerning control strategies applicable to the DEMO reactor.

Challenges towards an acceleration in stellarator reactors engineering: The dual coolant lithium–lead breeding blanket helical-axis advanced stellarator case

Under the ambitious EUROfusion mission of “bringing the stellarator line to technological maturity”, the development of a Dual Coolant Lithium–Lead (DCLL) Breeding Blanket (BB) for a Helical-Axis Advanced Stellarator (HELIAS) started. The BB is the crucial systems to achieve tritium self-sufficiency in fusion power plants. The DCLL has potentialities to answer the challenges posed by complex HELIAS configuration, having liquid breeder and decoupled first wall (FW) and breeding zone circuits. Both characteristics are being empowered to simplify remote maintenance and integration of the BB segments. Hence, a “quasi-toroidal” BB segmentation has been proposed to minimize the magnetohydrodynamic pressure drop eschewing complex electrical insulating components while hindering maintenance. Such controversy motivated the use of a detached FW small panels manageable through ports. On this basis, a decoupled Capillary Porous System FW has been assessed under thermal-hydraulics and neutronics point of view. Moreover, ad-hoc tools have been developed for the parametrization of models and convergence toward a viable DCLL HELIAS design, speeding-up the coupling of computer-aided design modelling and analyses. Such novel strategies are described in view of a potential acceleration of the stellarator reactor engineering and by providing an overview of the stellarator challenges inside a global context.

Embrittlement of WCLL blanket and its fracture mechanical assessment

In the European fusion programme, the Water Cooled Lithium Lead breeding blanket (WCLL BB) uses EUROFER as a structural material cooled with water at temperatures between 295 °C–328 °C and a pressure of 155 bar. The WCLL BB will be significantly irradiated (>2 dpa), while some parts will not receive significant heat loads, e.g. the sidewalls or the back-supporting structures. The irradiation, together with the irradiation temperature of EUROFER below 350 °C, produces a shift of the ductile-to-brittle-transition temperature (DBTT) to levels above room temperature at neutron doses, causing material damage as low as 2–3 dpa. Even though the DBTT does not reach the operating temperature level, brittle/non-ductile fracture is a concern during in-vessel maintenance when the BB temperature is below the DBTT. Two loading scenarios were identified as severe in this respect: (i) re-pressurization of the WCLL BB cooling loop after in-vessel maintenance, and (ii) dead weight loads during lifting of the BB segment. The embrittlement of the WCLL BB was investigated by quantifying the local DBTT shift in its parts based on current knowledge of the embrittlement behaviour of EUROFER under neutron irradiation. Therefore, a suitable, not overly conservative procedure was derived considering dpa damage and transmuted helium effects. The results demonstrate the ability to identify the 3D spread of the severely embrittled zones in the structure whose impact on the structural integrity was assessed considering the risk of brittle/non-ductile fracture. Thereby, the fracture mechanics approach established in nuclear codes was applied assuming its applicability to EUROFER. The embrittled zones in the first wall (FW) and its sidewalls pass the criteria when assessing the relatively low stresses resulting from the coolant pressure. The assessment was then continued considering stresses appearing in the FW during maintenance, in particular, when lifting the BB segment and transporting it out of the vacuum vessel. In this context, the maximum tolerable flaw sizes were determined in a parameter study considering designs of the FW with different cooling channel wall thicknesses.

Neutronic Assessments towards a Novel First Wall Design for a Stellarator Fusion Reactor with Dual Coolant Lithium Lead Breeding Blanket

The Stellarator Power Plant Studies Prospective R&D Work Package in the Eurofusion Programme was settled to bring the stellarator engineering to maturity, so that stellarators and particularly the HELIAS (HELical-axis Advanced Stellarator) configuration could be a possible alternative to tokamaks. However, its complex geometry makes designing a Breeding Blanket (BB) that fully satisfies the requirements for such a HELIAS configuration, which is a difficult task. Taking advantage of the acquired experience in BB design for DEMO tokamak, CIEMAT is leading the development of a Dual Coolant Lithium Lead (DCLL) BB for a HELIAS configuration. To answer the specific HELIAS challenges, new and advanced solutions have been proposed, such as the use of fully detached First Wall (FW) based on liquid metal Capillary Porous Systems (CPS). The proposed solutions have been studied in a simplified 1D model that can help to estimate the relative variations in Tritium Breeding Ratio (TBR) and displacement per atom (dpa) to verify their effectiveness in simplifying the BB integration and improving the machine availability while keeping the main BB nuclear functions (i.e., tritium breeding, heat extraction and shielding). This preliminary study demonstrates that the use of FW CPS would drastically reduce the radiation damage received by the blanket by 29% in some of the selected configurations along with a small decrease of 4.9% in TBR. This could even be improved to just a 3.8% TBR reduction by using a graphite reflector. Such an impact on the TBR is considered affordable, and the results presented, although preliminary in essence, have shown the existence of margins for further development of the FW CPS concept for HELIAS, as they have been not found, at least to date, to be significant showstoppers for the use of this technological solution.

Development of a thermal-hydraulic model of the EU-DEMO Water Cooled Lithium Lead Breeding Blanket Primary Heat Transport System

The EUROfusion consortium is developing the project of a DEMOnstration Fusion Reactor (EU-DEMO) which would follow ITER in the pathway towards the quest for the exploitation of fusion energy. EU-DEMO has been conceived to deliver net electric power to the grid. Therefore, proper critical evaluations of the tokamak cooling and power conversion systems are needed because they play a pivotal role in the design and licencing of the overall plant. The EU-DEMO reactor will be based on the tokamak concept and, as such, it is supposed to undergo a pulsed duty cycle under normal conditions, which might challenge the qualified lifetime of the main equipment inducing undue thermal and mechanical cycling. Moreover, the EU-DEMO plasma control strategy postulates the possible occurrence of planned and off-normal plasma overpower transients that might jeopardise the structural integrity of the plasma facing components. It is, therefore, of paramount importance to have appropriate tools to reproduce the thermal-hydraulic behaviour of tokamak cooling systems during major operational and accidental scenarios in a realistic and reliable way. In this context, University of Palermo in cooperation with EUROfusion has developed a finite volume model of the Primary Heat Transport System (PHTS) feeding the EU-DEMO Water Cooled Lithium Lead Breeding Blanket (WCLL BB). The activity has been led following a theoretical–computational approach based on the adoption of the TRACE thermal-hydraulic system code. Particular attention has been paid to capturing all the main geometrical, hydraulic and heat transfer features characterising both in-vessel and ex-vessel components. Preliminary analyses have also been carried out to check the code’s predictive potential in fusion relevant applications. Models, assumptions, and outcomes of the analyses are herewith reported and critically discussed.

Lithium-Lead/water interaction: LIFUS5/Mod3 series E tests analysed by SIMMER-III coupled with RELAP5

The Breeding Blanket is a necessary component to close the nuclear fusion reactor fuel cycle. amongst the most promising conceptual design, there is the Water Cooled Lithium Lead Breeding Blanket, with water as coolant and eutectic Lithium-Lead as neutron multiplier and breeder. The possible interaction between water and Lithium-Lead poses a main safety concern and prompted the scientific community to develop a numerical analysis tool able to simulate such a complex interaction. The SIMMER-III code was modified by UNIPI to simulate the chemical interaction between water and Lithium-Lead, furthermore also a coupling methodology between SIMMER-III and RELAP5/Mod3.3 was developed. The coupling tool employs SIMMER-III code to simulate the zone of Lithium-Lead interacting with water, whilst the RELAP5 code is used to simulate the water pipelines. The LIFUS5/Mod3 facility at the ENEA Brasimone Research Centre was designed to perform reliable experimental activities on the interaction between water and Lithium-Lead. In this facility water at high pressure is injected inside a reaction vessel, where the thermodynamic and chemical interaction between water and Lithium-Lead occurs. The experimental activities are divided in different tests matrix, the Series D and Series E tests. The two series differ in the amount of water injected during the transient. In series D the mass of water is predetermined whilst in series E water was injected continuously for a pre-fixed interval of time and the total injected mass was estimated a posteriori. This work shows the results of the coupling tool applied to Series E. The comparison between the experimental and numerical results is performed by identifying and characterizing the phenomena involved in the interaction. Furthermore, the overall performance of the coupled codes in the simulation of the phenomena is presented here.

Compatibility of niobium, vanadium and P22 steel in high temperature flowing LiPb

The LiPb loops of the WCLL BB (Water Cooled Lithium-Lead Breeding Blanket) have the primary aim of routing tritium-rich LiPb from the Breeding Blanket to the Tritium Extraction System of DEMO Reactor. Besides, the loops maintain LiPb in the desired conditions and purify it from corrosion and activated products and from helium generated by the neutron transmutation of lithium. For some of these tasks the loops need special functional materials. Among them, niobium and vanadium are widely considered for the tritium extraction technologies for their high permeability to tritium. In addition, the loops will be made of structural materials capable of withstanding the operative conditions and the aggressive environment. The current reference material for the piping is the ASTM A335 Grade P22. This paper investigates the compatibility at 500 °C of niobium, vanadium and P22 with LiPb flowing at 0.5 m/s. This velocity is the maximum foreseen in the LiPb loops, while the temperature was chosen following a conservative approach, as the maximum expected temperature is 450 °C in the BB and 330 °C in the LiPb loops. The specimens of the three materials were exposed for 1,000, 2,000 and 4,000 h in IELLLO LiPb facility at ENEA Brasimone Research centre. After the exposure, the specimens were examined by SEM-EDX, both in cross-section and surface view, XRD analyses were performed in order to investigate potential corrosion compounds formed. Weight loss analyses were also performed on niobium specimens, after chemical cleaning from adherent LiPb. The analyses highlighted the impact of corrosion products and their interaction with refractory metals (Ni and Fe) and P22 steel (Cr, after redeposition).

Assessment of SIMMER-III code in predicting Water Cooled Lithium Lead Breeding Blanket “in-box-Loss of Coolant Accident”

The in-box Loss of Coolant Accident is a major safety concern for the Water Cooled Lithium Lead Breeding Blanket design. SIMMER-III code has been modified by University of Pisa and ENEA C.R. Brasimone to perform deterministic safety investigation of such accidental scenario. Thanks these modifications, this version of the code has unique features for dealing with the PbLi water reaction related phenomena. The reliability of the code requires a systematic validation activity, carried out applying a standard methodology based on a three-steps procedure and through qualitative and quantitative evaluations. The methodology was applied to the LIFUS5 campaign. Post-test analyses highlighted open issues of test execution and of experimental data, as well as code limitations and capabilities. Nevertheless, the modified version of SIMMER-III code for fusion application is able to predict the relevant thermal-hydraulic phenomena during PbLi/water interaction.

Investigation of heat transfer in a steam generator bayonet tube for the development of PbLi technology for EU DEMO fusion reactor

In the frame of the EUROfusion roadmap for the development of the DEMO power plant, a research activity was carried out to develop a Lithium–Lead/water heat exchanger. The component should be capable to remove nuclear heat deposited in the liquid metal of the Dual Coolant Lithium Lead breeding blanket and feeding a steam turbine, ensuring an efficient thermal power conversion to electricity. One of the selected configurations is the steam generator bayonet tube. The HERO test section is an experimental mock-up in a relevant scale of this steam generator, consisting of a bundle of seven double-wall bayonet tubes with a leakage monitoring system. This test section, developed by ENEA at Brasimone Research Center and installed in the main vessel of the CIRCE pool facility, aims to investigate the thermal-hydraulic features of the system, providing a database for thermal-hydraulic system codes validation. An experimental campaign was carried out to demonstrate technological feasibility and performances of the prototypical heat exchanger, suitable as steam generator for the PbLi loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. A post-test analysis has been realized with RELAP5-3D and RELAP5/Mod3.3 codes in order to evaluate code capability in simulating heat transfer in liquid metal side.

Investigation of the DEMO WCLL Breeding Blanket Cooling Water Activation

Within the framework of the activities foreseen by the EUROfusion action on the cooling water activation assessment for a DEMO reactor equipped with a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), the University of Palermo is involved in the assessment of dose rates induced by the decay of nitrogen radioisotopes produced by water activation, nearby the main components (e.g. isolation valves) of both First Wall (FW) and Breeder Zone (BZ) cooling circuits. In particular, the aim of this work is to evaluate the spatial distribution of nitrogen isotopes (16N and 17N) in the WCLL BB cooling circuits. To this purpose, a coupled neutronic/fluid-dynamic problem is solved following a theoretical-numerical approach and adopting an integrated computational tool mainly relying on the use of MCNP6 and ANSYS CFX codes. The operative procedure adopted foresees the assessment of the production rate distributions of nitrogen isotopes within FW and BZ cooling channels and tubes by means of totally heterogeneous neutronic analyses. A fully 3-D fluid-dynamic approach is, then, used to compute the nitrogen isotope concentrations within the In-Vessel complex flow domain, while a 1-D lumped parameters approach is adopted to calculate their distribution along the Ex-Vessel BB Primary Heat Transfer System. The results obtained, herewith presented and critically discussed, provided the necessary data to perform dedicated neutronic and photonic transport analyses and, hence, to assess the dose rates in the aforementioned target locations.

Po-210 production in the European DEMO fusion reactor

The radionuclide inventory plays a central role in the safety of nuclear installations both during operation and their decommissioning. In nuclear fusion reactors using Pb–Li tritium breeding blankets, the undesired production of radiotoxic 210Po is still an unresolved safety issue. In this work, neutron transport calculations and inventory calculations are combined to predict the 210Po inventory in a DEMO fusion reactor using either a helium cooled lithium lead or a water cooled lithium lead breeding blanket. In order to guarantee that the environmental 210Po release associated with an ex-vessel leak-of-PbLi accident remains below the no-evacuation limit, the 210Po concentration in the Pb–Li should be kept below 1500 appt. It was found that no Pb–Li purification is required to keep the 210Po concentration in DEMO below this limit. However, in case the Pb–Li makes direct contact with water, more volatile Po-containing (oxy-)hydroxides could form. If these species increase the 210Po release rate by more than a factor of two, safety measures will be required. Therefore, 210Po generation in DEMO does not pose a hazard in case of a regular ex-vessel leak-of-PbLi accident, unless possibly in case the Pb–Li makes contact with water.

Parametric study of the influence of First Wall cooling water on the Water Cooled Lithium Lead Breeding Blanket nuclear response

In the framework of EUROfusion Work Package International Cooperation R&D activities, a close collaboration has been started among University of Palermo, ENEA Brasimone and ENEA Frascati for the development of the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept. In this context, a research campaign has been carried out at the University of Palermo in order to investigate the influence of First Wall (FW) cooling water configuration on the nuclear response of the WCLL BB under irradiation in EU-DEMO, in order to gain useful indications for the WCLL BB pre-conceptual designs.To this end, three-dimensional nuclear analyses have been performed by MCNP5 v. 1.6 Monte Carlo code. A semi-heterogeneous MCNP model of the WCLL BB “single module segment” concept has been used and its nuclear response has been assessed for different FW configurations, centring the attention, mainly, on global quantities as nuclear power deposition and tritium breeding ratio. The results obtained provided some information on the nuclear features of the WCLL BB concept, useful for its design optimisation. The outcomes of this parametric study are herein presented and critically discussed.

Design of a System for Hydrogen isotopes Injection into Lead-Lithium

In a fusion reactor based on liquid breeding blankets, tritium is generated due to the neutron irradiation of lithium based alloy, such as eutectic lead-lithium (PbLi). Then, tritium is extracted from the liquid metal by means of technologies that are presently under development such as the vacuum sieve tray, the permeation against vacuum or the gas liquid contactors. Nevertheless, for the experimental validation of these technologies at laboratory scale, hydrogen isotopes cannot be generated in situ in the liquid metal as in the breeding blanket. Hence, a system able to inject the gas in the flowing liquid at desired concentrations is required, avoiding the formation of bubbles as a consequence of the low solubility of hydrogen/deuterium in PbLi. The system should be capable of solubilizing the hydrogen to replicate as close as possible the conditions of a breeding blanket.A design of an injector based on a permeable membrane is here presented, being the driving force the gradient of concentrations existing between the two surfaces of the membrane. A hydrogen transport model for a tube-in-tube injector has been developed, showing that the injected hydrogen flux is proportional to the tube radius. However, the change on this parameter, that affects the velocity of the liquid and thus the mass transport coefficient, has opposite consequences on the rate of injection whose final impact relays on the properties of the employed membrane.An evaluation of the physical and geometrical aspects of a conceptual injector is depicted with the aim of optimizing the design to obtain adequate injection rate depending on the facility where the injector is installed. Finally, a conceptual design of a system to be implemented in an experimental PbLi loop for the validation of the permeation against vacuum technique for tritium extraction from PbLi is presented. The injector is based on a multi-tube component made of niobium able to inject hydrogen at the same rate as it is extracted under relevant conditions for a Dual Coolant Lithium Lead breeding blanket.

Nuclear analysis of the HCLL “Advanced-Plus” breeding blanket

In the frame of the European R&D effort to develop a fusion demonstrator (DEMO) a new concept of Helium Cooled Lithium Lead breeding blanket, called “Advanced-Plus” design, is studied to improve the tritium breeding. The aim of this paper is to present the nuclear analysis of this new HCLL design carried out with the TRIPOLI-4® Monte Carlo code and the Joint European Fission and Fusion nuclear data version 3.2 (JEFF-3.2). Nuclear quantities such as: neutron flux, tritium breeding ratio, nuclear heating, displacement damage and helium production are reported. Sensitivity analysis is also carried out to investigate the impact, in tritium breeding, of the next DEMO baseline with reduced outboard thickness and the advantage, in tritium production point of view, of the single module segment against the current multi-module segment option.

Investigation on the “advanced-plus” Helium Cooled Lithium Lead Breeding Blanket design concept for TBR enhancement regarding thermal and mechanical behavior

In the framework of the European “HORIZON 2020” research program, the EUROfusion Consortium develops a design of a fusion demonstrator (DEMO). The Breeding Blanket (BB) directly surrounding the plasma is a major component ensuring tritium self-sufficiency, shielding against neutrons from D-T plasmas and heat extraction for electricity conversion. CEA-Saclay, with the support of Wigner-RCP and IPP-CR is responsible of developing the Helium Cooled Lithium Lead (HCLL) BB concept. In order to enhance Tritium Breeding Ratio (TBR), an alternative HCLL BB design, called “Advanced-Plus” concept, has been investigated, with the aim to reduce the amount of steel, while ensuring good thermal and mechanical behavior. The thermal behaviors of the horizontal Stiffening Plates (hSPs) with various channels designs in terms of dimensions and layout have been parsed on simplified FE models for verifying and optimizing the HCLL Breeding Zone (BZ) before analyzing a HCLL BB ¼ model. Moreover, sensitive analyses have been carried out for testing various hSPs pitches, alternative Helium hydraulic schemes inside the hSPs as well as the detachable FW option in applying a zero Heat Flux (HF) on the FW. Furthermore, mechanical calculations have also been performed to assess the behavior of the module in faulted condition. The results obtained with the Cast3M-qualified-FEM code are presented and discussed.

Large-scale behavior of sandwich-like FCI components within the EU-DCLL operational conditions

The Dual Coolant Lithium Lead breeding blanket concept being developed by EUROfusion uses PbLi to breed tritium and to refrigerate the blanket. The location of electrically insulating Flow Channel Inserts (FCIs) as tubes floating inside the liquid is considered a good solution to minimize the magnetohydrodynamic (MHD) drop. A steel-alumina-steel sandwich-like FCI was here analyzed. Neutronic studies provide an estimation of the FCI impact on the tritium breeding performance of the blanket while its behavior under irradiation was estimated under representative DEMO pulsed operational scenario for activation and transmutation assessments. The generated flow velocity profile due to MHD simulations suggests the formation of high jets in the 2mm-gap near the side-walls. The mechanical stability and the insulation properties of the FCIs were tested after submitting commercial alumina ceramic tubes to several thermal cycles and thermal gradients between surfaces. The thermomechanical analysis anticipates tolerable stresses in the alumina part under a representative temperature distribution, whereas the restriction to freely expand originates important peak stresses in the internal steel jacket. Results indicate the moderate viability of these steel protected ceramic structures as insulating devices for FCI applications, although the advantages of introducing the naked ceramic tube have been also pointed out.

Thermo-hydraulic analysis of EU DEMO WCLL breeding blanket

In the frame of EUROfusion Work Package Breeding Blanket, thermal-hydraulic engineering analyses were carried out at ENEA Brasimone, in close cooperation with Sapienza University of Rome, to investigate thermal and fluid-dynamic behavior of the Water Cooled Lithium Lead Breeding Blanket, which is a candidate option for the European DEMO nuclear fusion reactor.The research activity focused on the equatorial elementary cell of an outboard segment with the aim of obtaining a temperature map of breeding blanket domains and to evaluate breeding zone and first wall systems cooling ability. A three-dimensional finite volume model of the breeding unit was developed, adopting the commercial CFD code ANSYS CFX 15.0. Steady state simulations were carried out considering a heat flux of 0.5 MW m−2, a radial power density distribution applied to all domains, and buoyancies forces activated in PbLi region. Results show that FW and BZ systems have good cooling efficiency, maintaining the maximum temperature in PbLi and solid structures well below 500 °C. The thermal-hydraulic results are discussed, highlighting open issues and suggesting pertinent modifications to DEMO WCLL coolant system layout aimed at optimizing the design.

Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

CFD analysis of WCLL BB PbLi manifold

ENEA CR Brasimone has developed the new design of the Water Cooled Lithium Lead Breeding Blanket (WCLL BB). In the new design Breeding Zone (BZ) water coolant flows in radial-toroidal direction, and PbLi flows in radial-poloidal direction; a gap between the Back Plate (BP) and the BZ constitutes the PbLi inlet manifold. The paper presents the CFD analysis of the WCLL BB PbLi inlet manifold, performed by ANSYS CFX-15 code. The objective of the analysis is to investigate the PbLi flow paths in the manifold region and to optimize the mass flow rate distribution in the BZ of the module. A preliminary analytical analysis is performed to evaluate pressure drop at orifices and to select the diameter of the orifices. CFD simulations are performed with the selected geometries. The optimal geometric configuration is a compromise between the need to have low PbLi velocity, to limit MHD issues and Eurofer corrosion, and to preserve the structural capability of the stiffening plates to withstand the overpressure conditions. Results demonstrate that the layout of the manifold and the size of the orifices is suitable to uniformly distribute the PbLi in the BZ, fulfilling the fluid velocity constraints. Specific MHD analyses are required to evaluate more realistic flow paths and to demonstrate the feasibility of the design.

Design of the helium cooled lithium lead breeding blanket in CEA: from TBM to DEMO

The helium cooled lithium lead (HCLL) blanket concept was originally developed in CEA at the beginning of 2000: it is one of the two European blanket concepts to be tested in ITER in the form of a test blanket module (TBM) and one of the four blanket concepts currently being considered for the DEMOnstration reactor that will follow ITER. The TBM is a highly optimized component for the ITER environment that will provide crucial information for the development of the DEMO blanket, but its design needs to be adapted to the DEMO reactor. With respect to the TBM design, reduction of the steel content in the breeding zone (BZ) is sought in order to maximize tritium breeding reactions. Different options are being studied, with the potential of reaching tritium breeding ratio (TBR) values up to 1.21. At the same time, the design of the back supporting structure (BSS), which is a DEMO specific component that has to support the blanket modules inside the vacuum vessel (VV), is ongoing with the aim of maximizing the shielding power and minimizing pumping power. This implies a re-engineering of the modules’ attachment system. Design changes however, will have an impact on the manufacturing and assembly sequences that are being developed for the HCLL-TBM. Due to the differences in joint configurations, thicknesses to be welded, heat dissipation and the various technical constraints related to the accessibility of the welding tools and implementation of non-destructive examination (NDE), the manufacturing procedure should be adapted and optimized for DEMO design. Laser welding instead of TIG could be an option to reduce distortions. The time-of-flight diffraction (TOFD) technique is being investigated for NDE. Finally, essential information expected from the HCLL-TBM program that will be needed to finalize the DEMO design is discussed.