Blanket Module Research Articles

Measurement of tritium production in the helium cooled pebble bed test blanket module mock-up at JET during DTE2

Quite often, detectors for measuring nuclear performance and radiation quantities of relevance in fusion experiments are requested to withstand harsh working conditions due to intense neutron and gamma radiation fields. High temperature constitutes a further harsh element in some locations of the machine, where it is necessary to perform some on-line measurements, as expected in the breeding blanket. This is an essential component in future fusion power plants to provide tritium self-sufficiency and its performance must be continuously monitored. Some Test Blanket Modules (TBMs) will be installed in ITER to provide the first experimental data to validate the predictions on tritium production and recovery. In the meantime, within EUROfusion program, the mock-up of the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM), previously used for the TBM experiment at the Frascati Neutron Generator (FNG), had been installed at JET to test some detectors and for benchmarking numerical codes used for breeding blanket assessment during DTE2 campaign. A diamond detector, calibrated to measure the tritium production through neutron detection inside the HCPB TBM mock-up, was tested during some plasma pulses of the DTE2 campaign at JET. The main outcome is that, as far as neutron emission rate is below 1015 s−1, neutrons are properly detected along the plasma discharge evolution by TBM diamond detector, consistently with the JET neutron monitor KN1. Moreover, the amount of tritium measured (E) is 1.40 × 10–12 tritons per source neutron and the comparison with MCNP radiation transport simulation (C) gives a ratio C/E = 0.77. Such measurements, considered promising, and their comparison with calculations are discussed in the present work. Criticalities emerged are analyzed and some improvements proposed with the main purpose of speeding up signal processing to make the system capable of working at higher plasma neutron emission rates.

Effect of different heat inputs on the microstructure and mechanical properties of the laser welded joint of CLF-1 steel

Abstract In this study, we examined the effects of various heat inputs on the microstructure and mechanical properties of the laser-welded joints of CLF-1 low-activation ferritic/martensitic steel medium-thick plates, which are used in the Test Blanket Module (TBM). The results indicate that under four different welding heat inputs (2727 J/cm, 3000 J/cm, 3563 J/cm, and 4500 J/cm), the weld microstructure is primarily composed of a large amount of lath martensite and a small amount of δ-Fe. As the heat input increases, the average width of lath martensite in the weld microstructure progressively increases, while the average volume fraction and width size of δ-Fe experience a significant increase. When the heat input surpasses 3563 J/cm, δ-Fe exhibits a dense aggregation. The tensile properties of the joints created under different heat inputs are superior to those of the base material. At the lowest heat input, the maximum impact toughness of the weld reaches 40 J.

ITER materials irradiation within the D–T neutron environment at JET: post-irradiation radioactivity analysis following the DTE2 experimental campaign

Abstract This work presents the results following the first irradiation of ITER materials samples in a tokamak D--T plasma environment operating at significant fusion power. The materials exposed to this nuclear environment at the Joint European Torus during the DTE2 experimental campaign that took place in 2021 include representative ITER samples from various components such as poloidal field (PF) coil jacket samples, toroidal field coil radial closure plate steels, EUROFER 97 steel, W and CuCrZr materials from the divertor, Inconel-718, CuCrZr, and 316L stainless steel for blanket modules, as well as vacuum vessel forging samples.

The experimental results discussed include high-resolution gamma spectrometry measurements and analysis conducted with the post-irradiated samples, of which there were 68 in total. These samples were exposed through different experimental campaigns, including deuterium, deuterium--tritium and tritium phases. Supporting the analysis were 25 dosimetry foil-based neutron diagnostics and two 'VERDI' neutron spectrometry diagnostics. A further 12 samples for positron annihilation spectroscopy (PALS) were also irradiated. The irradiation of all these samples took place in a long-term irradiation assembly located near the JET vacuum vessel.

The post-irradiation analysis of the ITER material samples has yielded valuable insights into their material activation levels and radiation fields. Comparative assessments between experimental measurements and comprehensive neutronics simulations have demonstrated a significant level of agreement in this work, while also revealing some discrepancies in specific material instances. The data and interpretation from this work not only serve as a robust experimental foundation for enhancing the precision and predictability of neutronics simulation approaches for ITER and next-step devices but also present some opportunities for the refinement of simulation methodologies. In light of these findings, a series of recommendations have been proposed, aimed at improving confidence in nuclear predictions associated with materials that have been exposed to fusion nuclear environments and advancing understanding in this important domain.

Antifriction characteristics of the magnetron sputtered MoS2 coating of ITER blanket module bolted joints in high vacuum at elevated temperature [formula omitted] =250 °C

The high values of bolt preload force Fa = 90–1200 kN created by the ITER remote handling system can be achieved by using dry coating lubricants such as magnetron sputtering film MoS2 or WS2. An expression was determined to estimate the contact pressure based on the number thread turns and the width of the thread contact area in fastened assemblies that used the Spiralock technology. This study investigated the tribological aspects of the magnetron sputtering MoS2 coating at different thicknesses hc = 6.4–21 μm, contact pressures pa = 20–1200 MPa against the spherical pin (austenitic stainless steel 316L(N)-IG) at elevated temperatures of T = 250 °C in vacuum pv = 10−2–10−3 Pa. Friction data for “humid” and “dry” rectangular specimens with MoS2 coating were obtained and compared. With an increase in the thickness of the coating from 6.4 μm to 21 μm, we obtained a distinct stabilization the friction coefficient of the low friction coating for each given regime of wear that diminished from 0.09 ± 0.010 to 0.05 ± 0.005. The wear mechanism of the “layered MoS2 coating/metal substrate” tribosystem resulted in significant plastic deformation along the sliding direction of the MoS2 solid lubricant as well as microgrooving and microploughing of the coating by the asperities of the flat contact spot of the spherical pin. Regardless of the type of bolted pair alloys used, we concluded that it is necessary to deposit a low friction coating on both the internal and external threads of the ITER blanket module bolted joints that use the Spiralock technology.

Blistering Behavior of Beryllium and Beryllium Alloy under High-Dose Helium Ion Irradiation.

Beryllium (Be) has been selected as the solid neutron multiplier material for a tritium breeding blanket module in ITER, which is also the primary option of the Chinese TBM program. But the irradiation swelling of beryllium is severe under high temperature, high irradiation damage and high doses of transmutation-induced helium. Advanced neutron multipliers with high stability at high temperature are desired for the demonstration power plant (DEMO) reactors and the China Fusion Engineering Test Reactor (CFETR). Beryllium alloys mainly composed of Be12M (M is W or Ti) phase were fabricated by HIP, which has a high melting point and high beryllium content. Beryllium and beryllide (Be12Ti and Be12W) samples were irradiated by helium ion with 30 keV and 1 × 1018 cm-2 at RT. The microstructures of Be, Be12Ti and Be12W samples were analyzed by SEM and TEM before and after ion irradiation. The average size of the first blistering on the surface of Be-W alloy is about 0.8 μm, and that of secondary blistering is about 79 nm. The surface blistering rates of the beryllium and beryllide samples were also compared. These results may provide a preliminary experimental basis for evaluating the irradiation swelling resistance of beryllium alloy.

A methodology for the selection of low friction/anti-seize coating in applying bolted joints of ITER blanket modules

The application of bolted joints with the Spiralock®/Selflock® technology for remote handling system in fusion machines leads to develop a methodology for selection a low friction/anti-seize coating in applying sliding surfaces of high-loaded threaded fasteners. This study discusses the analytical engineering method for calculation the maximum/minimum allowable friction coefficients in the thread and on the bearing surface during the tightening/untightening procedure. Calculations were made to determine the maximum/minimum allowable coefficient of friction in threaded pairs used for the ITER Blanket Module Connections and enhanced flux First Wall Panel fasteners (bolts with a regular metric thread/nuts with a self-locking thread, Spiralock®/Selflock® technology) of the ITER fusion experimental reactor under maximum torque conditions of the robot manipulator (remote handling system) and tensile load (preload) generated in the bolt during the tightening/untightening procedure. Design features of the Spiralock® technology were derived for calculation the working ranges of thread friction coefficient in the threaded pairs of the Blanket Module Connection and enhanced Heat Flux First Wall Panels fasteners at different coefficient of thread proportionality values (ζ=0.45–1.00). The lowest thread friction coefficients were obtained by the coefficient of thread proportionality ζ=1.00: 0.022≤μth≤0.171 for the M24 × 3 SB Electrical strap central bolts, 0.013≤μth≤0.047 for the М52 × 4 Outboard Flexible Cartridge central bolts, 0.010≤μth≤0.055 for the M64 × 4 Inboard Flexible Cartridge central bolts. The new method of determining the total tightening torque of the M24 × 3, M52 × 4, M64 × 4 bolted joints fastened to the Spiralock® female thread under conditions of elastoplastic, plastic deformation was applied to assess the suitability the solid lubricating coating MoS2 (VNIINP-212) under preload force Fa=10–1200 kN. The performed design verification indicated the significant differences in the calculated tightening torque obtained according to the standard and proposed method for the Spiralock® female thread by the plastic deformation of the tip of bolt thread and change of sliding friction coefficient depending on the contact pressure.

Tritium transport modelling in Europe: Current status, multi-physics integration and proposal of material property data

The capability to predict tritium retention and permeation by accurate and reliable tritium transport models for the breeding blanket design development is one of the most relevant objectives of the European Test Blanket Module (TBM) Programme. This also represents a key aspect for the development and licensing of next-future fusion systems and power plants, including the European DEMO. Tritium analyses can be performed using either system-level models to calculate global blanket and the related ancillary system performances or detailed 3D models for spatially limited areas and critical regions: the choice between these two modelling approaches depend on the objective of each analyse. When looking at the accuracy of tritium transport simulations, two main aspects have to be considered: the first one is the completeness of the description of the physical phenomena participating to the tritium transport dynamics; the second one is the reliability of the tritium transport material property values used in the simulation activities.In this context, this paper has the objective to carry out a critical analysis on the status of the integration of the most relevant multi-physics elements, providing recommendations on the approach to follow, with a particular focus on the system-level codes. Moreover, a set of data and correlations on tritium transport materials properties is proposed as a first step towards a material property database to be shared within the breeding blanket community.

Low cycle fatigue of EUROFER97 investigated for test blanket modules of ITER: An interlaboratory study

Several reduced-activation ferritic-martensitic (RAFM) steel grades have been proposed for structural applications in fusion reactors since the 1990s by several countries. The European reference RAFM steel EUROFER97 has been produced in four batches since 1998. It is intended for fabrication of Test Blanket Modules (TBM) for ITER as well as the blanket first wall and divertor cassettes of DEMO.RCC-MRx is the design code developed for high-temperature, research and fusion reactors by AFCEN. Its latest edition contains a provisional section dedicated to EUROFER97, aggregating properties of the first two batches, whereas a full qualification and final inclusion of a material requires three batches minimum. The completion of the code is required for the design assessment of TBM units in the frame of the nuclear licensing process.The EUROfusion project coordinates efforts to broaden the knowledge of EUROFER97 properties, which will be used for closing the database gaps in RCC-MRx. Low cycle fatigue (LCF) tests are a significant part of the EUROfusion test programme, since the material fatigue properties are indispensable for the component assessment owing to the cyclic nature of a tokamak operation.In this work we report the results of interlaboratory LCF tests on EUROFER97 batch 3 performed in a temperature range relevant for its application in ITER (RT–550 °C), wide selection of total strain ranges (0.3–1.5 %) and two values of strain ratio Rε (0 and –1). Along with the estimated fatigue life and cyclic behaviour, a fractographic analysis done with scanning electron microscopy (SEM) is presented.

Test blanket modules piping systems calculation method according to RCC-MRx code

The ITER Test Blanket Modules (TBMs) are installed and operated inside the vacuum vessel (VV) at equatorial ports located within port plugs (PP). The Pipe Forest (PF) is a network of pipes connecting the Ancillary Equipment Unit (AEU) to the TBMs. Depending of the TBM, the Pipe Forest includes various piping system such as Water Coolant (350°, 188 bars), Helium Coolant (500°, 100 bars) Lithium Lead and Tritium Extraction System. During operating states, the Pipe Forest main function is to withstand loads in normal, incidental and accidental situations according to RCC-MRx code criteria (including design rules for mechanical components for fusion installations). It includes severe thermomechanical loadings due to the thermal expansion added to seismic and vertical displacements events. In addition, the PF layout must satisfy constructability in constraint environments and maximize the space available for man access, which constitutes a real challenge for design.Since no existing RCC-MRx post-processing tool exists off the shelves, a specific methodology for pipe stress analysis is being developed by ITER, CEA and ASSYSTEM. This methodology starts with a post-processing calculation tool of stresses and reaction forces for looping back into the structural calculation. Moreover, a specific extraction module from 3D model to processing tool allows the conversion of CAD data to calculation entry files. Finally, a feedback loop is also re-integrated into the core CAD model through an import module, enabling the tracking of displacements and evaluate the impact of incorporating pipe movements into the analysis.This paper describes the pipe stress analysis methodology and its application for Pipe Forest system design life cycle. It covers the historical evolution of the routing including introduction of expansion loops. More recently, it incorporates the integrated calculation of the entire port cell, spanning from the back of the TBM sets, up to the connection pipes of the vertical shaft.

Macroscopic transport characteristics in packed bed with a multi-physics inversion strategy for solid breeding blanket of HCCB TBM

The solid breeding blanket with typical packed bed structure is a key component for tritium breeding and thermal conversion in future fusion reactors. The pore scale model for multi-physical field solution of solid breeding blanket is fine but very expensive. The simplified packed-bed scale model is more favored in engineering especially for the system-level calculation of the blanket, but it needs to be inputted appropriate macroscopic transport parameters. In this work, the multi-physics calculations of blanket in Chinese HCCB TBM (Helium-Cooled Ceramic-Breeder Test Blanket Module) at pore scale and packed-bed scale were performed. It was found that the packed-bed scale model with traditional macroscopic transport parameters is not accurate enough, especially the prediction deviation of tritium concentration distribution could reach about 25 %. Hence, a novel multi-physics inversion strategy was proposed to predict the macroscopic transport parameters of the packed bed under the multi-field coupling conditions. The inversion results show that the convection of helium significantly enhances the heat and tritium diffusion in the packed bed, and the effective thermal conductivity and effective diffusivity could be increased by 18 % and 46 %, respectively. Then based on the results by multi-physics inversion strategy, the modified effective thermal conductivity and effective diffusivity correlation models for solid breeding blanket packed bed are constructed. With the modified correlation models, the relative prediction deviations of pressure drop, temperature distribution and concentration distribution are within 2.5 %, 0.6 % and 10 % respectively at different inlet velocities (0.05 m/s ∼ 0.25 m/s) and different inlet temperatures (300 K ∼ 900 K). The study in this paper would assist the thermal-hydraulics analysis of the solid breeding blanket in the fusion reactor.

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

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

RELAP5/Mod3.3 MHD module development and validation: WCLL-TBM mock-up model

Magnetohydrodynamic (MHD) phenomena significantly impact the design of magnetic-confinement fusion reactors, particularly in the context of breeding blankets (BB) utilizing liquid metals (LMs) as working fluids. These effects arise from the interaction between the electro-conductive flowing metal and the magnetic field used for plasma confinement in the reactor chamber. Induced electrical currents in the liquid generate Lorentz forces, thereby altering flow behaviour in comparison to standard hydrodynamic conditions. For example, MHD effects modify velocity distribution and mass transport within ducts, amplify pressure losses, and influence heat transfer mechanisms. Accurate estimation of these impacts is crucial for the effective design of a liquid metal breeding blanket. While computational tools are essential for fusion-related physical analyses, no comprehensive MHD code currently exists for simulating all relevant phenomena in a liquid metal blanket. In this context, models predicting both distributed and concentrated MHD pressure drops have been integrated into the thermal-hydraulic system code RELAP5/Mod3.3. The Verification and Validation (V&V) process compares code results to direct numerical simulations and experimental data. For validation, RELAP5 recreates experimental results of a Water Cooled Lithium Lead test blanket module at magnetic field intensities ranging from Ha=500−3000, confirming the reliability of the newly implemented MHD subroutines for predicting pressure drops within this parameter range.

Investigation of hydrogen generation from the reaction of beryllium pebbles with water vapor for WCCB-TBM

Be pebble, as a neutron multiplier material, is being explored for the development of a Water-Cooled Ceramic Breeder Test Blanket Module (WCCB-TBM) in Japan for the International Thermonuclear Experiment Reactor (ITER) project. The Be pebbles were fabricated by the rotating electrode process (REP). However, as the safety of the test blanket module (TBM) could be compromised by hydrogen generation via Be–H2O reaction owing to the leakage of pressurized water from the cooling pipes inside TBM, the safety assessment of TBM must be conducted using the Be–H2O reactivity data for the Be pebbles whose fabrication method is the same as that for TBM (i.e., REP). Thus, in this study, a Be–H2O reactivity test was conducted using Be pebbles that were fabricated by REP to characterize the generated hydrogen and evaluate the hydrogen generation (HG) rates toward WCCB-TBM safety assessment. First, the Be–H2O reactivity data were obtained at ≥ 300 °C (the coolant temperature for the TBM). The experimental results and oxidation rate of the Be pebbles revealed that only the Be surface may have been oxidized at 300 °C–600 °C, although its inside was oxidized at 700 °C–1000 °C. Further, the experimental data–derived HG rates at 300 °C–1000 °C increased with the temperature and the tendency changed in the three temperature regions. The HG rates did not contradict those reported by previous studies. Additionally, these regions corresponded to each expected Be–H2O reaction behavior at 300 °C–1000 °C. At a Be temperature of < 600 °C, the generated hydrogen from TBM will not exceed 2.5 kg of a limitation by the water leakage on the estimation based on the HG rates and the qualitative effects of the safety functions.

Instrumentation integration in the European Test Blanket Modules

Amongst the four different Test Blanket Modules (TBM) that shall be tested in ITER, the European domestic agency (Fusion For Energy) participates in two concepts: the Water-Cooled Lithium Lead (WCLL) and the Helium-Cooled Ceramic Pebbles (HCCP). A progressive set of activities including implementation and integration of instrumentation has been conducted from preliminary design stage. The current work is mainly focused on the thermal and mechanical instrumentation, although other sensors are also considered. The integration of instrumentation in the TBMs is challenging; consequently, the number and position of the sensors should be carefully optimized. An important advance of the current work is the use of reconstruction algorithms to provide an estimation of the reconstruction ability of the different sensor sets. These algorithms comprise a series of mathematical tools which use model results and sensor measurements to infer different variables at positions where no in-situ measurement is available. Apart from the use of these algorithms, the final decision on sensor location has been conducted from a holistic point of view considering the different perspectives and integration challenges.

Assessment of the relevancy of ENEA Water Loop facility with respect to ITER WCLL TBS Water Cooling System by considering their thermal-hydraulic performances

The Water-Cooled Lead-Lithium (WCLL) is one of the two candidate concepts for the Breeding Blanket (BB) of DEMO. A Test Blanket Module (TBM) together with its Water Cooling System (WCS) is going to be installed and tested in the ITER reactor. The WCS acts as primary cooling circuit of the TBM module, and it is designed to reproduce the water thermodynamic conditions expected at the DEMO BB inlet.During last years, ENEA and the DIAEE of Sapienza University of Rome have carried out the conceptualization of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone. The W-HYDRA platform is composed by three individual facilities called: Water Loop, Steam, and LIFUS5/Mod4. Water Loop replicates the salient thermal-hydraulic features of the ITER WCLL WCS, and it is equipped with a test section placed inside a Vacuum Vessel (VV) to investigate mock-ups of the whole TBM or its individual parts.This paper assesses the relevancy of the WL facility with respect to ITER WCLL TBM System comparing their thermal-hydraulic performances during selected operational and accidental conditions. Two RELAP5/Mod3.3 models were developed, and the outcomes of the simulations showed a good agreement, thus demonstrating the effectiveness of WL facility to support the ITER TBM program.

Design and Development of Hydrogen Isotopes Extraction System at IPR

Tritium breeding, recovery and safe storage are very important for achieving self-sustainability in a nuclear fusion reactor. Indeed, it is one of the essential task in terms of safety and fuel cycle aspects. Usually, tritium breeding is performed in the blanket module. In Indian blanket module, there are mainly two types of tritium breeder materials, one is Lithium based ceramic pebbles Li2TiO3 (Lithium Titanate) as solid breeders and second is the lithium based alloy Pb-Li (Lead Lithium) as a liquid breeder. Tritium bred in solid breeders is extracted with purge helium gas, whereas tritium bred in liquid Pb-Li needs suitable gas-liquid contactor. It should be noted that, tritium recovery (i.e.,-extraction) from Pb-Li is very challenging due to low solubility of tritium in liquid Pb-Li and no such separation technology readily available commercially. Therefore, the necessity for developing a reliable and efficient tritium recovery technology (i.e.,- extractor) for liquid Pb-Li is of key interest.In present work, design and development of Hydrogen Isotopes Extraction System (HIES) for hydrogen recovery from liquid Pb-Li is discussed. As hydrogen and tritium are isotopes, hydrogen is used instead of tritium in present system due to several safety concerns with tritium. Recovery of hydrogen from liquid Pb-Li is carried out using structured packing as a Gas-Liquid Contactor (GLC). The design details of gas-liquid contactor (i.e.-extractor column) and various process tanks of HIES are discussed. In present design, configuration of vertically mounted extractor column makes overall HIES relatively tall (∼ 3 m height) and use of liquid Pb-Li as process fluid makes the system heavy weight (∼1500 kg) too. In addition to this, system needs to be operated above Pb-Li melting temperature (> 235 C) to keep Pb-Li in liquid form. Therefore, thermal analysis of piping as well as seismic analysis of whole system is carried out and discussed along with various operational stages of HIES.

Development of water-cooled cylindrical blanket in JA DEMO

The concept of the tritium breeding blanket for Japan’s DEMOnstration fusion reactor (JA DEMO) has been developed with pressure tightness against in-box loss-of-coolant accidents based on a water-cooled solid breeder concept. The cooling conditions are designed on the pressurized-water reactor water conditions which are the coolant temperature of 290 °C–325 °C and the operating pressure of 15.5 MPa, respectively. The point of the blanket design is to reduce the amount of structural material in casing as well as to ensure its pressure tightness. This is because a decrease in the amount of structural material improves tritium breeding ratio (TBR). A cylindrical structure, a thin wall casing structure which ensures pressure tightness and could increase TBR, is feasible. However, a relatively larger useless space is expected between modules when the cylindrical blanket modules are arranged in a vacuum vessel, which could decrease TBR. Therefore, the cylindrical blanket modules are to be in a close-packed arrangement to reduce useless space. The Be12Ti block (which shows a minor swelling compared to Be) is selected to achieve the target TBR as the net Be density is equivalent to the case of the Be pebble. The use of Be12Ti blocks can reduce or remove the cooling piping inside the module as the Be12Ti block has a higher thermal conductivity than pebbles. As a result of the neutronics, finite element method, and computational fluid dynamics analyses, it was found that the target TBR value can be achieved in cylindrical structure blanket that ensure pressure tightness.

Overview of progress on water cooled ceramic breeder blanket in Japan

National institutes for Quantum Science and Technology (QST) are the implementing body of ITER project in Japan and takes the lead in fusion DEMO reactor development in Japan. The primary candidate of the breeder blanket for JA DEMO is Water Cooled Ceramic Breeder (WCCB) blanket. ITER-Test Blanket Module (TBM) program is an opportunity for the demonstration of DEMO breeding blanket concept under actual fusion environment. QST has led the development of the WCCB TBM Concept in the program since the signing of the TBM Arrangement in 2014. The preliminary design of test blanket system (TBS) is on-going. Recently, the status of the preliminary design of TBS was assessed between QST and the ITER Organization (IO). So, the status of the latest TBS design and the DEMO relevancy is presented.

An integral methodological framework for the thermo-mechanical analysis and structural integrity assessment of the European TBM sets

Two European Test Blanket Modules (TBM) concepts are being developed by F4E, the Water-Cooled Lithium Lead (WCLL) and the Helium-Cooled Ceramic Pebbles concept (HCCP, in collaboration with ITER KOREA). For both concepts, the ESTEYCO Mechanics team is responsible for designing and analyzing the TBM-box, which serves as the plasma-facing module for tritium breeding, and the TBM Shield in the WCLL concept. The design-by-analysis activities for these TBM-Sets are particularly challenging due to several factors. These include diverse and complex loads such as thermal, electromagnetic, pressure, and seismic loads, intricate failure modes to be assessed, the use of a new structural material (EUROFER 97) with limited available information, and the need to work within tight compliance margins in a restricted design space. Over the years, ESTEYCO has developed advanced methodologies to virtually test various design solutions systematically, coherently, and efficiently, aligning with the applicable nuclear design code (RCC-MRx). This paper provides an overview of key developments in simulating the TBM-Set under a comprehensive set of loads and subsequently assessing its structural integrity in an exceptionally detailed and comprehensive manner. The authors have created novel automated tools that encompass the entire process. These tools address tasks like distributing nuclear thermal loads, simulating cooling system behavior under operational conditions, characterizing and distributing electromagnetic loads, implementing code assessment rules for numerous evaluation points, and developing robust visualization tools for result analysis and validation.

Major improvements in the WCLL TBM design towards next review gates

The European domestic agency (Fusion for Energy) is developing the WCLL-TBS (Water Cooled Lithium Lead Test Blanket System) concept for its implementation and testing in ITER. The system is based on the use of the eutectic Pb–15.7Li enriched at 90 % in 6Li as tritium breeder and tritium carrier. Water is used as coolant with current target nominal inlet/outlet temperatures of 295/328 °C and 15.5 MPa pressure (similar to PWR conditions). The Test Blanket Module –TBM– is the in-vessel component where tritium generation will take place and whose structural material is EUROFER 97. Its design has been significantly improved in the last years. These design improvements have focused on three major areas: (i) overcoming some weaknesses of the previous design in the rear manifold area associated to the stiffening rod concept, (ii) strengthening the link with manufacturing developments and (iii) optimizing the amount of EUROFER 97 in the TBM box, so that the design converges to the limits imposed by the ITER project requirements. The work presented in this paper describes the latest design evolutions achieved in the WCLL TBM design with respect to that presented at CDR and, in particular, the major re-design of the TBM in the manifold area based on the ‘crossing vertical’ stiffening plate concept proposed and developed by the ESTEYCO Mechanics design team in collaboration with F4E.