Lithium Lead Research Articles

Steady state thermo-mechanics and material property definition framework for analyzing DCLL blanket in the fusion nuclear science facility

A thermo-mechanics model that relies on creating the material property definition framework (MPDF) and multiphysics coupling of the heat transfer and the solid mechanics modules is developed to determine the structural integrity of the recently designed dual cooled lead lithium (DCLL) inboard blanket (IB) for the Fusion Nuclear Science Facility under steady state loads. The MPDF is called to supply fusion relevant neutron irradiation and temperature induced changes in material properties during multiphysics finite element runs, and PbLi temperature profiles are used to approximate Magnetohydrodynamics effect and the nuclear volumetric heating on the PbLi. Neutron irradiation and temperature induced reduction of the yield and ultimate strengths of F82H steel at the first wall (FW) are quantified for one year. A blanket in an assembly with gaps between blanket sectors and another blanket in an assembly with no gaps between blanket sectors, both exposed to radiation damage that lasted for one year are analyzed. Analysis using the elastic ITER structural design criteria for in-vessel components (ITER SDC-IC) design rules and a linear isotropic-hardening-type elastoplastic material model are used where most appropriate. The IB blanket with gaps between blanket sectors will withstand the steady state combined thermal and coolant loads for one year operational period but will fail if no gaps are allowed between blanket sectors. It is recommended that a gap of about 7.62 mm should be provided between IB blanket sectors during assembly which would close up during service, stop neutron streaming, reduce stresses and reduce bending of the FW into the scrape-off layer.

The Source Permeator System and Tritium Transport in the TEX PbLi Loop

Permeation is investigated for the introduction of hydrogen isotopes into lead lithium (PbLi) for the Tritium Extraction eXperiment (TEX). TEX is a forced-convection PbLi loop under construction at Idaho National Laboratory that will test the vacuum permeator (VP) method of tritium extraction from PbLi. The source permeator (SP) delivers atomic hydrogen (H, D, and T) from a gas-phase reservoir into the PbLi via a permeable dense metal membrane. A modular system and a fixed SP system are presented. In the modular design, PbLi flows through the inside of a tubular membrane, and gas-phase hydrogen is introduced on the outside of the membrane. Atomic hydrogen permeates radially inward through the membrane into the PbLi. In the fixed design, PbLi flows into an expansion chamber with closed-ended tubular membranes inserted. Gas-phase hydrogen is introduced on the inside of the closed-ended membranes, and atomic hydrogen permeates radially outward into the flowing PbLi. Hydrogen transport models based on steady-state mass transport through PbLi and permeation through the metal membrane were developed to assess the operation of the SP relative to experimental variables and to allow understanding of uncertain parameter effects, such as PbLi hydrogen transport properties and the effective hydrogen permeability of the VP. This modeling effort considers iron as the SP material and vanadium as the VP material.

Dual-Resistance of Ion Migration and Moisture Erosion via Hydrolytic Crosslinking of Siloxane Functionalized Poly(Ionic Liquids) for Efficient and Stable Perovskite Solar Cells

Dual-Resistance of Ion Migration and Moisture Erosion via Hydrolytic Crosslinking of Siloxane Functionalized Poly(Ionic Liquids) for Efficient and Stable Perovskite Solar Cells

The solubility of helium in lead–lithium eutectic alloy

• The He solubility value needed to evaluate helium saturation conditions and bubble nucleation in nuclear fusion liquid metal blankets. • He solubility values in lead and lead–lithium eutectic proposed by using semi-empiric correlations sustained on cohesion models and historic data for Li, Na, K and Hg. • Satisfactory soundness of semi-empirical correlation based on Kumaŕs cohesion model to approach available historic measurements. • New experimental approaches needed to confirm the low values obtained. The helium production rates in liquid metals (Pb-Li eutectic alloy, LLE and others) Blanket Breeders (BB) are nearly mol-to-mol linked to tritium and intimately associated with the compulsory requirement of high tritium self-sufficiency of next fusion reactors. When LLE is oversaturated and the helium solubility limit is exceeded, helium atoms can nucleate in the form of bubbles. The presence of helium bubbles within LLE channels could have severe impact on the diverse BB designs, in particular on tritium transport permeation and its recovery. Even though He is an inert gas assumed to be insoluble, the helium Henry’s constant ( K H ) in a liquid metal is not zero. The very low K H and difficulties to measure it has historically driven to the absence of basic data. A semi-empirical correlation is proposed providing the helium solubility (i.e.: the helium Henry’s constant) based on Kumar’s cohesion model using the available thermo-physical experimental solubility data for lithium, sodium, potassium, mercury. The proposed expression for eutectic lead–lithium is: K H at . fr . Pa - 1 = 5.99 · 10 - 4 exp ( - 1.16 · δ k ) exp - 54.23 / RT , being K H the Henry’s constant; T [K], R [8.314 10 -3 kJ mol −1 K −1 ] and δ k the Kumar’s cohesive parameter. From a δ k justified value of 18.2 MPa 1/2 in LLE the values for K H range from 1.14·10 -17 to 1.35·10 -15 at.fr.Pa −1 for temperatures between 350 and 870C. The helium solubility should integrate the lead–lithium eutectic nuclear material database for fusion systems design.

Design and Micro-Fabrication of Focused High-Frequency Needle Transducers for Medical Imaging.

In this study, we report an advanced fabrication technique to develop a miniature focused needle transducer. Two different types of high-frequency (100 MHz) transducers were fabricated using the lead magnesium niobate-lead titanate (PMN-0.3PT) and lithium niobate (LiNbO3) single crystals. In order to enhance the transducer’s performance, a unique mass–spring matching layer technique was adopted, in which gold and parylene play the roles of the mass layer and spring layer, respectively. The PMN-0.3PT transducer had a 103 MHz center frequency with a −6 dB bandwidth of 52%, and a signal-to-noise ratio (SNR) of 42 dB. The center frequency, −6 dB bandwidth, and SNR of the LiNbO3 transducer were 105 MHz, 66%, and 44 dB, respectively. In order to compare and evaluate the transducers’ performances, an ultrasonic biomicroscopy (UBM) imaging on the fish eye was performed. The results showed that the LiNbO3 transducer had a better contrast resolution compared to the PMN-0.3PT transducer. The fabricated transducer showed an excellent performance with high-resolution corneal epithelium imaging of the experimental fish eye. These interesting findings are useful for the future biomedical implementation of the fabricated transducers in the field of high-resolution ultrasound imaging and diagnosis purpose.

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

The experimental qualification of the tritium extraction unit (TEU) from the LiPb eutectic alloy (15.7 at. % Li), the breeder material of the water-cooled lithium–lead breeding blanket concept, is one of the fundamental items for the demonstration of tritium balance sustainability for ITER and DEMO fusion reactors. Several technologies have been proposed as TEU, but the selection of the reference technology can be carried out only after the experimental measurement of the tritium extraction efficiency. For this purpose, a dedicated facility, called TRIEX-II, was designed and installed at ENEA Brasimone Research Centre, Italy. The facility is able to qualify gas–liquid contactor (GLC), permeator against vacuum and liquid-vacuum contactor technologies at different temperatures, lithium–lead mass flow rates and hydrogen isotopes concentrations. In TRIEX-II, the hydrogen or deuterium, used to simulate tritium, are solubilised inside the LiPb with a dedicated saturator and are then extracted from the liquid metal in the GLC mock-up, which uses pure helium or a mixture constituted by helium and hydrogen as stripping gas and works in the temperature range between 300 and 500 °C.

Maturation of critical technologies for the DEMO balance of plant systems

The Pre-Concept Design (PCD) of the Balance of Plant (BoP) systems of the EU-DEMO power plant is described in this paper for both breeding blanket (BB) concepts under assessment, namely the Water Cooled Lithium Lead (WCLL) BB and the Helium Cooled Pebble Bed (HCPB) BB. Moreover, the results of a preliminary evaluation of a number of BoP variants are discussed.This paper outlines the steps of the BoP design development, highlighting the project objectives and the strategy for their achievement under the very challenging requirements which include, among others, the intermittent nature of the DEMO plasma heat source.The main achievements during the PCD Phase will be reported together with the development plan for the Concept Design (CD) Phase to reach a mature (feasible) BoP concept for DEMO.

Production and transport modelling of Po-210 in DEMO reactor

One of the generic designs of the nuclear fusion DEMO reactor proposed by the EUROfusion consortium foresees the development of a tritium breeding blanket (BB) relying on the use of the liquid-metal PbLi eutectic alloy as both neutron multiplier and tritium breeder, namely the water-cooled lithium lead (WCLL) BB, whose strengths and weaknesses are well known. This paper focuses the attention on one of the possible disadvantages of such a technology: the production of the highly radiotoxic radionuclide 210Po, which could become a safety issue to be accounted for. The 210Po concentration within the PbLi circuit has been assessed by solving a modified version of Bateman’s equations to consider the alloy circulation, so a one-dimensional convective fluid-dynamic model has been set up. Nuclear quantities have been evaluated by Monte Carlo neutron transport analyses using MCNP code and adopting a fully heterogeneous model of DEMO equipped with the WCLL BB. Moreover, rough sensitivity analyses have been performed to assess the influence on the results of the uncertainties related to the 209Bi radiative-capture cross section and the initial concentration of this nuclide which is present in the PbLi as an impurity. Results obtained have been critically discussed and some safety issues have been addressed to evaluate the possible hazard in case of a leak of PbLi accident.

Establishing technical specifications for PbLi eutectic alloy analysis and its relevance in fusion applications

One of the key aspects on the future operation of liquid metal-based breeding blankets is the eutectic composition of the lead–lithium (PbLi) alloy. There is discrepancy on the exact eutectic point, varying the lithium content from 15 at% to 17 at%, which will be discussed in this work. Additionally, the PbLi to be used during the R&D activities requires a careful characterization. A guide that compiles a detailed description of the material handling, analytical techniques to be applied and results interpretation is presented for an improved use of this alloy in experimental fusion applications. Some of the employed techniques are thermogravimetry, inductively coupled plasma mass spectrometry and X-ray diffraction, among others. In CIEMAT, PbLi samples from different providers have been analyzed showing a lithium dispersion from 3 at% to 18 at% and with an impurity content up to 400 ppm.The exact lithium content and the presence of impurities have a high impact on the experimental activities in support to PbLi related technologies. A crucial point is how it affects neutronic calculations on the Tritium Breeding Ratio. In addition, it is reported a bibliographic review denoting the existence of a large experimental uncertainty in the solubility values for hydrogen isotopes in PbLi. It is known the higher solubility of hydrogen in lithium against lead, therefore small deviations on the lithium percentage may imply an inaccuracy on those values. This reasoning can be also applied on the experimental determination of the diffusivity values.

Design of the Test Section for the Experimental Validation of Antipermeation and Corrosion Barriers for WCLL BB

Tritium permeation into the Primary Heat Transfer System (PHTS) of DEMO and ITER reactors is one of the challenging issues to be solved in order to demonstrate the feasibility of nuclear fusion power plants construction. Several technologies were investigated as antipermeation and corrosion barriers to reduce the tritium permeation flux from the breeder into the PHTS. Within this frame, alumina coating manufactured by Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) are two of the main candidates for the Water Cooled Lithium Lead (WCLL) Breeder Blanket (BB). In order to validate the performance of the coatings on relevant WCLL BB geometries, a mock-up was designed and will be characterized in an experimental facility operating with flowing lithium-lead, called TRIEX-II. The present work aims to illustrate the preliminary engineering design of a WCLL BB mock-up in order to deeply investigate permeation of hydrogen isotopes through PHTS water pipes. The permeation tests are planned in the temperature range between 330 and 500 °C, with hydrogen and deuterium partial pressure in the range of 1–1000 Pa. The hydrogen isotopes transport analysis carried out for the design and integration of the mock-up in TRIEX-II facility is also shown.

Preliminary MHD pressure drop analysis for the prototypical WCLL TBM with RELAP5/MOD3.3

In fusion reactor development, the design of the breeding blanket (BB) represents an important technical challenge due to the extreme operative conditions that the component must withstand. One of the critical issues of a liquid metal BB is the occurring of magnetohydrodynamic (MHD) effects within the piping network that transports the electro-conductive fluid. Predicting MHD phenomena is fundamental for a reliable and efficient blanket design and computational tools are indispensable to achieve a fusion-relevant physical analysis. For this reason, a dedicated MHD module for the System Thermal-Hydraulic (STH) code RELAP5/MOD3.3 (RELAP5) is under development at Sapienza University of Rome. In its current state, RELAP5 can evaluate MHD pressure losses in many common layouts for BB piping system. In a previous work of the Authors, the verification and validation (V&V) procedure of the MHD module has been discussed. In this paper, the code is used to assess the MHD pressure loss in the Water-Cooled Lithium Lead Test Blanket Module (TBM). Furthermore, the geometry of the TBM box is optimized to reduce the MHD pressure loss and to equalize the flow distribution among the Breeding Units.

Thermo-Mechanical Analysis and Design Update of the Top Cap Region of the DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment

Within the framework of the EUROfusion research and development activities, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) is one of the two candidates to be chosen as the driver blanket for the European DEMO nuclear fusion reactor. Hence, an intense research work is currently ongoing throughout the EU to develop a robust conceptual design able to fulfil the design requirements selected at the end of the DEMO pre-conceptual design phase. In this work, the thermo-mechanical analysis and the design update of the top cap (TC) region of the DEMO WCLL Central Out-board Blanket (COB) segment is presented. The scope of the work is to find a design solution of the WCLL COB TC region able to fulfil the design requirements, prescribed by the reference RCC-MRx code, under the selected nominal and accidental steady state loading scenarios. The activity herein presented moved from the WCLL COB reference design, purposely modified in compliance with the adopted thermal and mechanical requirements in order to attain a robust TC region geometric layout. In the end, a geometric configuration called “TC region-mod++” was determined, foreseeing a TC able to safely withstand both nominal and accidental loads. Nevertheless, some criticalities still hold in the internal stiffening plates and, therefore, further and finer analysis are necessary to fully match the goal. In any case, it was also found that the proposed approach for the design update is promising and worthy to be further pursued. The work was performed following a theoretical–numerical approach based on the finite element method (FEM) and adopting the quoted Ansys commercial FEM code.

Analysis of the Thermo-Mechanical Behaviour of the EU DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment under an Optimized Thermal Field

Within the framework of the EUROfusion research activities on the DEMO Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) design, a research study was performed to preliminarily optimize, from the thermal point of view, the WCLL Central Outboard Blanket (COB) segment in order to investigate its structural behaviour under a realistic thermal field. In particular, a study of thermal analyses was performed to optimize the Double Walled Tubes and Segment Box cooling channels’ geometric configurations along the poloidal extension of the WCLL COB segment, in order to obtain a spatial temperature distribution fulfilling the thermal design requirement. Then, the thermo-mechanical analysis of the WCLL COB segment under Normal Operation (NO, representing nominal conditions), Upper Vertical Displacements Event (UVDE, representing a plasma disruption event) and Over-Pressurization (OP, representing an in-box loss of coolant accident) scenarios were carried out, assuming the previously obtained thermal field, to realistically predict displacement and stress fields. Finally, a stress linearization procedure allowed comparing the stress values obtained in some critical regions of the structure with the criteria prescribed by the reference design standard RCC-MRx. A theoretical–numerical approach based on the Finite Element Method (FEM) was followed using the commercial code Abaqus v. 6.14.

Toward full simulations for a liquid metal blanket: part 2. Computations of MHD flows with volumetric heating for a PbLi blanket prototype at Ha ∼ 104 and Gr ∼ 1012

On the pathway toward full simulations for a liquid metal (LM) blanket, this part 2 extends a previous study of purely magnetohydrodynamic (MHD) flows in a DCLL blanket in reference Chen et al (2020 Nucl. Fusion 60 076003) to more general conditions when the MHD flow is coupled with heat transfer. The simulated prototypic blanket module includes all components of a real LM blanket system, such as supply ducts, inlet and outlet manifolds, multiple poloidal ducts and a U-turn zone. Volumetric heating generated by fusion neutrons is added to simulate thermal effects in the flowing lead–lithium (PbLi) breeder. The MHD flow equations and the energy equation are solved with a DNS-type finite-volume code ‘MHD-UCAS’ on a very fine mesh of 470 × 106 cells. The applied magnetic field is 5 T (Hartmann number Ha ∼ 104), the PbLi velocity in the poloidal ducts is 10 cm s−1 (Reynolds number Re ∼ 105), whereas the maximum volumetric heating is 30 MW m−3 (Grashof number Gr ∼ 1012). Four cases have been simulated, including forced- and mixed-convection flows, and either an electrically conducting or insulating blanket structure. Various comparisons are made between the four computed cases and also against the purely MHD flows computed earlier in reference Chen et al (2020 Nucl. Fusion 60 076003) with regards to the (1) MHD pressure drop, (2) flow balancing, (3) temperature field, (4) flows in particular blanket components, and (5) 3D and turbulent flow effects. The strongest buoyancy effects were found in the poloidal ducts. In the electrically non-conducting blanket, the buoyancy forces lead to significant modifications of the flow structure, such as formation of reverse flows, whereas their effect on the MHD pressure drop is relatively small. In the electrically conducting blanket case, the buoyancy effects on the flow and MHD pressure drop are almost negligible.

Structural Assessment of the European DEMO Water-Cooled Lithium Lead Breeding Blanket Central Outboard Segment Under Remote Maintenance Loading Conditions

Within the framework of the European research activities devoted to design the DEMO breeding blanket (BB), the Water-Cooled Lithium-Lead (WCLL) BB concept is one of the candidates currently assessed in EU. One of the main issues connected to its design is represented by the remote maintenance (RM) operations. In particular, previous thermo-mechanical analysis highlighted a Ductile to Brittle Transition Temperature (DBTT) shift above the room temperature within the WCLL central outboard (COB) segment during normal operation. This may pose problems during the RM operations foreseen in the BB replacement phase. Indeed, within those WCLL COB segment’s regions where the DBTT shifted above the room temperature, the stress level arising during RM operations may be able to promote crack growth and migration resulting in a fragile fracture event. Then, in this work, the structural assessment of the DEMO WCLL COB segment under RM scenarios is performed, in order to provide the fracture mechanics analysis with the necessary input in terms of stress spatial distribution and to understand if the deformation occurring during RM can be accommodated by the available space reservation. Several potential RM scenarios have been considered, taking into account also the potential occurrence of a seismic event or the necessity to remove the WCLL COB segment with the liquid breeder not drained. The study has been performed adopting a numerical approach based on the Finite Element Method and adopting the quoted Abaqus code. The obtained results are quite encouraging and will allow, in the follow up, the refinement of the RM plan foreseen for the WCLL BB.

Spectral and Transmittance Properties of Er3+ Doped Zinc Lithium Lead Calcium Borophosphate Glasses

Abstract: Zinc lithium lead calcium borophosphate glasses containing Er3+ in (40- x):P2O5:10ZnO:10Li2O:10PbO:10CaO:20B2O3:xEr2O3 (where x=1, 1.5,2 mol %) have been prepared by melt-quenching method. The amorphous nature of the glasses was confirmed by x-ray diffraction studies. Optical absorption, Excitation, fluorescence and Transmittance spectra were recorded at room temperature for all glass samples. Judd-Ofelt intensity parameters Ωλ (λ=2, 4, 6) are evaluated from the intensities of various absorption bands of optical absorption spectra. Using these intensity parameters various radiative properties like spontaneous emission probability, branching ratio, radiative life time and stimulated emission cross–section of various emission lines have been evaluated. Keywords: ZLLCBP Glasses, Optical Properties, Judd-OfeltTheory, Transmittance Properties.

The synergy of heavy-ion irradiation and lithium-lead corrosion on deuterium permeation behavior of ceramic coating

Functional ceramic coatings have been investigated for several decades to mitigate tritium permeation through structural materials in a fusion reactor blanket. For the establishment of a liquid blanket system, the coatings require not only tritium permeation reduction but also the tolerance to irradiation damage induced by neutrons and corrosion by liquid tritium breeders. In this study, deuterium permeation measurements under exposure to liquid lithium–lead were performed for a heavy-ion-irradiated zirconium oxide coating sample to elucidate the synergy of irradiation and corrosion on hydrogen isotope permeation. The irradiated coating decreased the deuterium permeation flux under lithium–lead exposure by more than two orders of magnitude in comparison with the unirradiated one. Each activation energy of permeation and diffusion of the coating sample was larger than that of the unirradiated one at around 450 °C and decreased to the level of the unirradiated one at 600 °C. The results indicated that the voids formed by ion irradiation aggregated in grain boundaries of the coating, resulting in an increase in the energy barrier of diffusion at low temperatures and then a decrease due to the defect recovery at high temperatures. Besides, a corrosion product layer formed on the coating during lithium–lead exposure increased the energy barrier of solution. The synergy of irradiation and corrosion on permeation was developed at lower temperatures before the defect recovery.

Activities in Divertor Reflector and Linear Plates Using WCLL and HCPB Breeding Blanket Concepts

In fusion devices, such as European Demonstration Fusion Power Reactor (EU DEMO), primary neutrons can cause material activation due to the interaction between the source particles and the targeting material. Subsequently, the reactor’s inner components become activated. For safety and safe performance purposes, it is necessary to evaluate neutron-induced activities. Activities results from divertor reflector and liner plates are presented in this work. The purpose of liner shielding plates is to protect the vacuum vessel and magnet coils from neutrons. As for reflector plates, the function is to shield the cooling components under plasma-facing components from alpha particles, thermal effects, and impurities. Plates are made of Eurofer with a 3 mm layer of tungsten, while the water is used for cooling purposes. The calculations were performed using two EU DEMO MCNP (Monte Carlo N-Particles) models with different breeding blanket configurations: helium-cooled pebble bed (HCPB) and water-cooled lithium lead (WCLL). The TENDL–2017 nuclear data library has been used for activation reactions cross-sections and nuclear reactions. Activation calculations were performed using the FISPACT-II code at the end of irradiation for cooling times of 0 s–1000 years. Radionuclide analysis of divertor liner and reflector plates is also presented in this paper. The main radionuclides, with at least 1% contribution to the total value of activation characteristics, were identified for the previously mentioned cooling times.

The DEMO Water-Cooled Lead–Lithium Breeding Blanket: Design Status at the End of the Pre-Conceptual Design Phase

The Water-Cooled Lead–Lithium Breeding Blanket (WCLL BB) is one of the two blanket concept candidates to become the driver blanket of the EU-DEMO reactor. The design was enacted with a holistic approach. The influence that neutronics, thermal-hydraulics (TH), thermo-mechanics (TM) and magneto-hydro-dynamics (MHD) may have on the design were considered at the same time. This new approach allowed for the design team to create a WCLL BB layout that is able to comply with different foreseen requirements in terms of integration, tritium self-sufficiency, and TH and TM needs. In this paper, the rationale behind the design choices and the main characteristics of the WCLL BB needed for the EU-DEMO are reported and discussed. Finally, the main achievements reached during the pre-conceptual design phase and some remaining open issues to be further investigated in the upcoming conceptual design phase are reported as well.

Multiphysics Optimization for First Wall Design Enhancement in Water-Cooled Breeding Blankets

The commercial feasibility of the first fusion power plant generation adopting D-T plasma is strongly dependent upon the self-sustainability in terms of tritium fuelling. Within such a kind of reactor, the component selected to house the tritium breeding reactions is the breeding blanket, which is further assigned to heat power removal and radiation shielding functions. As a consequence of both its role and position, the breeding blanket is heavily exposed to both surface and volumetric heat loads and, hence, its design requires a typical multiphysics approach, from the neutronics to the thermo-mechanics. During last years, a great deal of effort has been put in the optimization of the breeding blanket design, with the aim of maximizing the tritium breeding and heat removal performances without undermining its structural integrity. In this paper, a derivative-free optimization method named “Complex method” is applied for the design optimization of the European DEMO Water-Cooled Lithium Lead breeding blanket concept. To this purpose, a potential performances-based objective function, focusing on the maximization of the tritium breeding, is defined and a multiphysics numerical model of the blanket is developed in order to solve the coupled thermo-mechanical problem, while the optimization algorithm leads the design towards a minimum optimum point compliant with the prescribed requirements. Once the optimized design is obtained, its nuclear and thermo-structural performances are assessed by means of specific neutron transport and multiphysics simulations, respectively. Finally, the structural integrity is verified by means of the application of the RCC-MRx design criteria. • A tool for First Wall optimization in water-cooled breeding blankets is developed. • The optimization is performed using the gradient-free method named “Complex method”. • An objective function aiming at tritium breeding maximization is defined. • A FE model is used to check the compliance with the thermo-structural requirements. • The optimized design is finally assessed by proper, more detailed numerical analyses.