Technical Meeting on Tritium Breeding Blankets and Associated Neutronics

2 Sept 2025, 08:00 → 5 Sept 2025, 18:00 Europe/Vienna

IAEA Headquarters, Vienna, Austria

Richard Kamendje (EUROFusion)

30 June 2025	Deadline for submission of abstracts through IAEA-INDICO for regular contributions
30 June 2025	Deadline for submission of Participation Form (Form A), Form for Submission of a Paper (Form B) and Grant Application Form (Form C) (if applicable) through the official channels
30 July  2025	Notification of acceptance of abstracts and of assigned awards
02 September 2025	Event begins
05 September 2025	Event ends

For their economical viability, fusion power plants based on a D-T fuel cycle will have to breed their own tritium in-situ. Yet, tritium breeding blankets are one of the most important and novel technical components in any planned next step devices to follow ITER. Many nations are embarking on dedicated R&D programmes to select tritum breeding blanket concepts which overcome issues, such as feasibility and attractiveness. 

Against this backdrop, the IAEA is launching a series of Technical Meetings on the topic of tritium breeding blankets and associated neutronics.

The objectives of the first Technical Meeting in the series are threefold: 

• Provide a forum for exchanging and discussing technical issues associated to the development and nuclear qualification of tritium breeding blanket technologies;

• Contribute to the establishment of specific guidelines for the qualification of tritium breeding blankets as part of fusion nuclear installations;

• Contribute to harvesting the return on experience from the ITER TBM Program to inform and support the development of breeding blankets for next step fusion facilities. 

 

***

A special issue in the journal Fusion Engineering and Design on “Progress in tritium breeding blanket development and nuclear qualification” is being considered to publish contributions presented at this meeting following a due peer review process.

Authors of accepted contributions at the meeting are kindly invited to consider submitting their manuscript for publication in this special issue. Information regarding publication modalities including timeline and deadline for manuscript submission will be provided as soon as available. In any event, it is expected that the submission of manuscripts will start immediately after the end of the meeting.

 

Infosheet_TM_Tritium_Breeding_02_05_September_2025_final.pdf

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Forms A, B and C

• Form A_EVT2304553.docx
• Form B_EVT2304553.docx
• Form C_EVT2304553.docx

Tuesday 2 September

09:00 Arrival

09:30 Welcome and Opening Remarks

Introduction by IAEA and Chair of the meeting

Topic I: Session 1: Chaired by Richard Kamendje

Conveners: Gabriele Ferrero (Politecnico di Torino), Kazunari katayama (Kyushu University), Lei Chen (Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP)), Luciano Giancarli (ITER Organisation), Qiang Qi (Institute of Plasma Physics, Chinese Academy of Sciences)

1 The ITER TBM Program: A Pathway Towards Tritium Breeding Blankets For D-T Power Reactors

In ITER, two dedicated equatorial ports are allocated for the TBM (Test Blanket Module) Program in order to allow the operation and testing of Test Blanket Systems (TBS) that are relevant mock-ups of tritium breeding blankets systems for fusion power reactors, using operation parameters typical for power reactors. The ITER baseline 2024 foresees two phases, DT-1 and DT-2. During DT-1, the plan is to operate four TBSs [1], namely:
— TBS-1: water-cooled lithium-lead TBS procured by EU.
— TBS-2: helium-cooled ceramic pebble TBS procured jointly by Korea and EU.
— TBS-3: water-cooled ceramic breeder TBS procured by Japan.
— TBS-4: helium-cooled ceramic breeder TBS procured by China.

Each TBS is formed by an in-vessel component where tritium is generated, called TBM, and by several sub-systems (e.g., coolant system, tritium-extraction system, measurement systems). The four TBMs are inside the equatorial ports # 16 and #18, directly facing the plasma (two TBMs in each port) and the corresponding subsystems are located behind the ports in specific areas of the Tokamak Complex (see Fig.1).

Other types of TBSs could be operated during DT-2, for instance a dual-coolant lithium-lead type, and other TBM designs using different structural and functional materials.

After a general description of potential breeding blanket systems and of the selected TBSs for DT-1, the presentation addresses some examples of on-going R&D and the associated technical information that can be obtained in support of DEMO breeding blankets development, such as, for instance, the development of the Reduced-Activation Ferritic/Martensitic (RAFM) steels as TBM structural materials, the development of codes and standards, the approach to pressure equipment implementation and to the tritium management, the development of specific measurements and control systems. These technical outcomes can be obtained already in the on-going TBS design phase to be completed by 2029 and continue during the manufacturing phase, until installation planned to start in 2035. The four DT-1 TBSs will start operation in 2039 and will operate for about ten years.

It will also be shown that TBS operations during DT-1 will allow to achieve the main top-level testing objectives for each TBS, provided that the ITER operational scenarios implement some minimal operational requirements specific for the TBM Program. Starting from the top-level testing objectives, several measurable “general Campaign Testing Objectives (gCTO)” have been identified. They cover each of the five DT-1 campaigns and are applicable to all TBSs. The achievement of these objectives will give useful information also for other types of TBSs and the corresponding results would be a useful support to the specific modelling activities that are under development for the characterization of tritium breeding blankets for DEMO reactors.

Speaker: Luciano Giancarli (ITER Organisation)

2 Design and Multi-physical Performance Analysis of the WCCB and COOL Blankets for CFETR

The water-cooled ceramic breeder (WCCB) blanket has been developing as a near-term solid blanket candidate for the Chinese Fusion Engineering Testing Reactor (CFETR). Meanwhile, the supercritical CO2 cOoled Lithium-Lead (COOL) blanket has been proposed as an advanced candidate in recent years. This presentation reports the overall design and performance analysis for both WCCB and COOL blankets.
The WCCB blanket features a mixed pebble bed of Li2TiO3 and Be12Ti as the tritium breeder and neutron multiplier, and a pressurized water of 15.5 MPa as the coolant. The feasibility of the WCCB blanket design is evaluated from the aspects of neutronics, thermo-hydraulics, thermal-mechanics, tritium breeding and nuclear safety. 3D neutronics analysis shows TBR is 1.123 when considering the port effect and contribution from divertor blankets, and the blanket can provide enough shielding capacity. Thermo-mechanical analysis indicates that the WCCB blanket are compliant with the material temperatures and stress limit. Safety analysis for typical blanket module #3 proves that temperature and pressure vary in allowable ranges under steady, transient, LOFA and in-vessel LOCA condition. Tritium transport analysis of typical blanket module #3 provided the amount of tritium inventory and permeation. Tritium permeation into coolant is necessary to be extracted by coolant purification system and reduced by tritium permeation barrier.
The COOL blanket is a typical dual coolant liquid blanket, namely PbLi of 460-600°C for cooling breeding zones and supercritical CO2 at 350–400 °C for cooling FW and structures. Addition-ally, the electrically and thermally insulating SiCf/SiC composites are utilized as Flow Channel Inserts (FCIs) to isolate the high-temperature corrosive PbLi and mitigate the magnetohydrodynamic (MHD) effect. Similarly, comprehensive analyses concerning neutronics, thermomechanics, thermal hydraulics, MHD, and safety demonstrate the blanket concept's feasibility. A 3D neutronic analysis shows that the current design can achieve a global TBR of 1.166 in case of port effects and divertor blankets. In addi-tion, thermal hydraulic analysis indicates the thermally insulating FCI is necessary to prevent the struc-tural overheating. Furthermore, thermomechanical analyses for typical blanket slices present allowable temperature and stress on structures even under the PbLi outlet of 700 °C. Moreover, MHD analysis proves utilizing electrically insulating FCIs in the PbLi inlet channel decreases the pressure drop dra-matically from 0.75 MPa to 0.09 MPa.
In order to test the breeding blanket technology, both WCCB and COOL Test Blanket Modules (TBM) will be manufactured and configured in two middle ports of the Burning Plasma Experimental Superconducting Tokamak (BEST), a DT operating fusion device constructed in the next five years. Design and test plan of both TBMs will be briefly introduced in the end.

Speaker: Lei Chen (Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP))

11:10 Coffee break

3 The Performance Evaluation of Tritium Breeding Materials Under Service Conditions

In D-T fusion reactors, tritium self-sufficiency is one of the key factors for maintaining steady-state operation. The key factors of tritium breeding materials are effective tritium breeding and stability of performance under long-term service. In terms of tritium breeding, a high lithium atom density is essential to produce substantial amounts of tritium, and good tritium release properties are required for efficient tritium recovery. However, the tritium release process is highly complex, influenced by various factors such as grain size, porosity, radiation defects, purge gas, and surface condition. Regarding stability over long-term service, it is crucial to ensure the stability of mechanical properties, thermal conductivity, and compatibility of tritium breeding materials under harsh operating conditions, namely high temperatures and high-flux neutron irradiation. These properties are crucial for the engineering design, normal operation, and safety of tritium breeding blankets. These issues require comprehensive research and analysis. The China Fusion Engineering Test Reactor (CFETR) prefers solid breeder blankets, primarily comprising two blanket design schemes: a water-cooled ceramic blanket using Li2TiO3 as the tritium breeding material and a helium-cooled ceramic blanket using Li4SiO4. Based on the main issues faced by tritium breeding materials under service conditions, this work comprehensively analyzes the tritium release behavior and the stable performance of tritium breeding materials under service environments.
Tritium release performance was evaluated, and the influence of various factors on tritium release was investigated. Experimental results indicate that tritium retention diminishes with escalating temperatures, while microstructure attributes such as grain size, porosity exert an influence on tritium release. Long-term service at high temperatures with He + 0.1%H2 flowing gas, showed that the average crushing load of Li2TiO3 pebbles remained stable, whereas the average crushing load of Li4SiO4 pebbles decreased. Significant microstructural changes, including grain size, porosity, and lithium loss, were observed for Li4SiO4 pebbles. The failure behavior of tritium breeding materials under prolonged annealing conditions was studied using the Weibull distribution analysis method, providing failure probabilities at different stresses. Long-term exposure to high temperatures affects the thermal conductivity of the breeding materials, with Li2TiO3 remaining relatively stable while the thermal conductivity of Li4SiO4 decreases. High-temperature lithium evaporation and irradiation accelerate the corrosion of tritium breeding materials on structural materials.
This work comprehensively analyzes the tritium release behavior, thermal stability and compatibility of tritium breeding materials under service conditions, providing data support for the blanket design of CFETR.

Speaker: Qiang Qi (Institute of Plasma Physics, Chinese Academy of Sciences)

4 Performance Assessment and Integration of Different Tritium Extraction Technologies in the Outer Fuel Cycle

The outer fuel cycle (OFC) of a fusion power plant (FPP) is a fundamental component that allows retrieving the bred tritium to permit D-T plasma operations. Moreover, this system plays a fundamental role for the safety of workers and population, acting as a physical barrier that delimits the tritiated circuit, and extracts tritium from the system, thereby diminishing the tritium inventory both in the BB and in ancillary systems.
The open-source TRItium Object-oriented and Modular Analysis (TRIOMA) code is a Python package the analysis of OFCs in FPPs. The code leverages on pre-built classes and functions to build an intuitive, object-oriented framework. Its 0-dimensional analytical formulation enables fast calculation of the extraction efficiency and of integral parameters, including tritium inventory and losses, under steady-state conditions, for each component of the OFC, enabling rapid design iterations for OFC optimization. Both molten salt and liquid metal breeders, which have different analytical formulations due to different solubility laws [1], [2], [3], are included in TRIOMA. The code can be easily employed to make the preliminary design of an OFC, to analyze feasibility and sensitivity on input properties, to observe start-up transients, and pulsed plasma operations. Moreover, due to the flexibility of the Python environment, the integration with other open-source tritium transport tools and data analysis is facilitated.
The presented work employs TRIOMA to design and compare different tritium extraction technologies within the same OFC circuit. In particular, the comparison involves the Permeator Against Vacuum (PAV), the Gas Liquid Contactor (GLC) based on a packed tower column, and the Liquid Vacuum Contactor (LVC) technologies. These are the main selected tritium extraction technologies for PbLi liquid breeders foreseen in the WCLL breeding blanket module. In particular, the PAV mock-up experiment is based at ENEA Brasimone [4] in the TRIEX facility [5], as well as the GLC mock-up, while the LVC technology is tested in CIEMAT laboratories.
REFERENCES
[1] P. W. Humrickhouse and B. J. Merrill, “Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets,” Fusion Science and Technology, vol. 68, no. 2, pp. 295–302, Sep. 2015, doi: 10.13182/FST14-941.
[2] J. D. Rader, M. S. Greenwood, and P. W. Humrickhouse, “Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion,” Nuclear Technology, vol. 203, no. 1, pp. 58–65, Jul. 2018, doi: 10.1080/00295450.2018.1431505.
[3] C. Alberghi, L. Candido, M. Utili, and M. Zucchetti, “Development of new analytical tools for tritium transport modelling,” Fusion Engineering and Design, vol. 177, p. 113083, Apr. 2022, doi: 10.1016/j.fusengdes.2022.113083.
[4] F. Papa et al., “Engineering design of a Permeator Against Vacuum mock-up with niobium membrane,” Fusion Engineering and Design, vol. 166, p. 112313, 2021.
[5] M. Utili, C. Alberghi, L. Candido, F. Papa, M. Tarantino, and A. Venturini, “TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors,” Nucl. Fusion, vol. 62, no. 6, p. 066036, Apr. 2022, doi: 10.1088/1741-4326/ac5c74.

Speaker: Gabriele Ferrero (Politecnico di Torino)

5 Experiments and Modeling on Tritium Mass Transfer in Solid Breeding Blanket

From the viewpoint of fuel control and tritium safety in a DT fusion reactor, it is important to correctly understand the tritium behavior in the blanket system. Previous studies have shown that the tritium generated by the nuclear reaction of lithium and neutron is released as both of HT and HTO from solid breeder materials such as Li2TiO3 and Li4SiO4. The HT / HTO ratio depends on the kind of breeder material and the purge gas. Since the permeation rate of HT through the metal wall is orders of magnitude faster than that of HTO, the permeation rate into the cooling water through the cooling tubes differs depending on the HT / HTO ratio, which affects tritium recovery efficiency. Some of the tritium implanted into the first wall of the blanket is also incorporated to the cooling water via the diffusion in the metal walls such as tungsten and F82H. Since mass transfer phenomena on the plasma facing wall are complicated, and it is difficult in the present to accurately evaluate the tritium transfer rate toward to the cooling water in consideration of this phenomenon correctly. Nevertheless, it is important to work on modeling with the aim of reflecting the results of fundamental experiments and improving the accuracy of tritium behavior prediction. The tritium permeated from the blanket module to the cooling water is transferred to the secondary cooling water via the heat exchanger. For safety management of primary and secondary cooling water and power generation equipment, it is required to evaluate the tritium transfer rate from the blanket modules to the cooling water. In this work, based on the research activities so far, the characteristics of tritium behavior from the blanket to the secondary cooling water in DEMO conditions were summarized.

Speaker: Kazunari Katayama (Kyushu University)

13:10 Lunch

Topic I: Session 2: Chaired by Alessandro Spagnuolo

Conveners: David Pickersgill (United Kingdom Atomic Energy Authority), Davide Pettinari (Politecnico di Torino), Gaetano Bongiovi (Università degli Studi di Palermo), Dr Guangming Zhou (Karlsruhe Institute of Technology (KIT)), Hiroyasu Tanigawa (Japan Atomic Energy Agency)

6 Overview of JA DEMO Breeding Blanket Development

The water-cooled solid breeder concept has been adopted as the concept of tritium breeding blanket for Japanese DEMOnstration fusion reactor (JA DEMO). The coolant water conditions are similar to those of pressurized water reactor, with the temperature of 290 – 325°C and an operation pressure of 15.5 MPa. The design has evolved to simultaneously meet the requirements of structural integrity (including pressure tightness against in-box loss-of-coolant), achievement of the target tritium breeding ratio (TBR), and manufacturability. A cylindrical blanket design using beryllide (Be12Ti) in block form as a functional structural element with neutron multiplication and heat transfer has been proposed as a design concept that satisfies the above diverse requirement [1,2].
In this concept, nuclear heat from the tritium breeding area is removed through contact heat transfer between the beryllide blocks and the inner wall of the casing. A tritium breeder (Li2TiO3) pebbles are filled between the inner surface of the cylindrical structure and the outer surface of the beryllide block to ensure maximum tritium production. Thermal analysis considering contact heat transfer confirmed that the casing structure is designed to meet the operational temperature range of the tritium breeding material and structural materials. Thermomechanical analysis showed that the primary and secondary stress values of the enclosure structure are below the allowable stresses for all evaluation lines. Nuclear analysis of the local TBR indicated that the TBR is estimated to exceed 1.20.
A technical readiness assessment (TRA) was conducted to identify critical technology elements (CTEs) and technical readiness level (TRL) of JA DEMO BB [3]. CETs were identified in the area of structural material development, functional material development and system integration. The current status and strategies for the design methodology of irradiated structures, as well as the efforts to develop the material property handbook (MPH) of the structural materials (reduced activation ferritic/martensitic steel, F82H) and functional materials, (beryllide blocks, etc.) will be discussed.

REFERENCES
[1] Y. SOMEYA, H. TANIGAWA, Y. SAKAMOTO, Nucl. Fusion 54 (2024) 046025.
[2] Hisashi TANIGAWA, et al., Fusion Eng.Des. 136 (2018) 1221–1225.
[3] Hiroyasu TANIGAWA et al., “DEMO Breeding Blanket R&D in Japan: Its Progress, Issues, and Prospects for Development with TRL Criteria Definition “, The 30th Symposium on Fusion Engineering (SOFE2023), (2023)

Speaker: Hiroyasu Tanigawa (National Institutes for Quantum Science and Technology)

7 Neutronic And Thermo-Mechanical Analysis in Support of The Design of The Water-Cooled Lead Ceramic Breeder Breeding Blanket Concept for The EU DEMO

The design of the breeding blanket (BB) system represents one of the biggest challenges towards the accomplishment of the conceptual design of the DEMO fusion reactor, due to its pivotal role in the machine operations. In this regard, novel BB concepts have been recently emerging in Europe with the aim of developing BB configurations capable of surmounting the major criticalities emerged so far from the dedicated BB R&D programme carried out within the EUROfusion consortium. In this context, the present work focuses on the Water-cooled Lead Ceramic Breeder (WLCB) BB concept. It foresees sub-cooled pressurized water as coolant, solid lead as neutron multiplier and the well-known ceramic breeder in form of pebble bed as breeding material. The ongoing studies aims at demonstrating that such a combination of functional materials will allow maximizing the tritium breeding while keeping high efficiency in heat removal and structural performances compliant with the provisions of the adopted code&standards. Therefore, in order to attain a robust WLCB BB design, neutronic and thermo-mechanical assessments are currently ongoing at the University of Palermo, under the umbrella of EUROfusion. In particular, different WLCB BB architectures have been investigated under the nuclear, thermal and structural standpoints to compare their performances in view of the goals of the WLCB BB design activities.
From the neutronic point of view, a campaign of parametric nuclear analysis has been launched. In particular, assuming different tungsten layer thickness values, different breeder material composition and structural material for the breeding zone cassette (a dedicated structure devoted to house the breeder and the neutron multiplier), the best configuration in terms of tritium breeding ratio has been selected. Then, it has been possible to properly evaluate the impact of each selected parameter on the WLCB BB overall nuclear response.
From the thermo-mechanical standpoint, the outcomes of the afore mentioned neutronic analysis, in terms of nuclear deposited heat power density within the different structural and functional materials, have been assumed to investigate the thermal and structural behaviour of the Top Cap region of the WLCB BB Central Outboard Blanket segment. The study has allowed checking the fulfilment of the prescribed thermal requirements on the Eurofer steel maximum allowable temperature and to verify the compliance with the RCC-MRx structural design criteria. Results have allowed providing indications for the design improvements of such a singular region.
The activity herein presented has been performed adopting a numerical approach, using the MCNP code for the nuclear analysis and the Ansys Workbench calculations suite for the thermo-mechanical investigations.

Speaker: Gaetano Bongiovi (Università degli Studi di Palermo)

8 Design status of the European DEMO Helium Cooled Pebble Bed breeding blanket

Within the EUROfusion DEMO programme, the Helium Cooled Pebble Bed (HCPB) breeding blanket is being developed as a reactor-relevant breeding blanket for the European DEMO. The reference design of this concept employs pressurized helium (8 MPa) as the coolant, a lithium-ceramic pebble bed as the tritium breeder, and beryllide blocks (beryllium-based alloys) as neutron multipliers. Tritium extraction employs a helium purge gas with 100 Pa of hydrogen (partial pressure) at 8 MPa, enhancing tritium recovery through the formation of hydrogen isotopes. This configuration achieves a tritium breeding ratio (TBR) of 1.17—exceeding the TBR design target of 1.15 for tritium self-sufficiency—alongside high reliability and low tritium permeation. However, the use of beryllium poses challenges for commercial deployment due to its global supply constraints, toxicity, and limited reserves. To ensure a cost-effective and economically viable solution, alternative neutron multipliers such as lead (Pb) or Pb-based alloys are being explored.
In terms of manufacturing, the conventional approach for fabricating the first wall involves hot isostatic pressing (HIP) or Electrical Discharge Machining (EDM) to create cooling channels, followed by bending the wall into a U-shape. However, the feasibility of scaling this method to produce an 18-meter-long first wall at the DEMO scale remains uncertain. Alternative manufacturing techniques, such as casting and forging, are currently being investigated as promising solutions. Consequently, modifications to the HCPB architecture are also being explored. This work presents the design progress of the evolved helium-cooled pebble bed breeding blanket concept, emphasizing design simplification and cost-effective material selection, along with its nuclear, thermal-hydraulic, and thermo-mechanical performance assessments.

Speaker: Guangming Zhou (Karlsruhe Institute of Technology (KIT))

16:10 Coffee break

9 OpenMC-based Parametric Neutronic Assessment of Fusion Breeding Blankets for Compact D-T Reactor Configurations

The choice of breeding blanket configuration represents one of the most critical design aspects for the viability of future D-T fusion power plants. POLITO and Eni have investigated and compared different breeding blanket concepts by evaluating their neutronic performance under consistent reactor boundary conditions. The objective is to provide a systematic assessment of the impact of blanket design choices on key nuclear parameters relevant for tritium self-sufficiency, structural integrity, and component lifetime.
To achieve this, a fully open-source and reproducible workflow was developed based on the OpenMC Monte Carlo code, employing Constructive Solid Geometry (CSG) models to enable parametric analyses of breeding blanket configurations for compact fusion reactors. The use of CSG has been found effective during the parametric studies, even if it shows some limitations when more realistic CADs are used. The plasma chamber and external shielding dimensions are kept constant across all configurations, while varying the internal blanket design within the vacuum vessel according to the investigated configurations. Both liquid and solid blanket concepts were analyzed, specifically FLiBe, CLiF, WCLL and DCLL liquid blankets, as well as HCPB as solid breeder designs.
For each configuration, key neutronic metrics were evaluated, including tritium breeding ratio (TBR), deposited power density, neutron flux distributions, energy spectra, and material activation. Furthermore, preliminary estimations of neutron-induced structural damage (displacements per atom, dpa) were performed as a function of the blanket concept. Additionally, a performance assessment of the computational codes is presented, emphasizing the efficiency and scalability of different modeling approaches. Uncertainty quantification (UQ) is carried out to evaluate the impact of nuclear data uncertainties on simulation results, providing insights into how cross-section uncertainties propagate through the calculations and affect final design metrics.
This work supports early design decisions for fusion power plant-class reactors by enabling comparisons across breeding strategies. within an open and reproducible computational environment.

Speaker: Davide Pettinari (Politecnico di Torino)

10 Reliability Optimised Blanket Using Simulation & Test: A Novel Approach to Breeder Blanket Design

Many public and privately funded fusion reactors are due to begin operation around the middle of the 21st century, the vast majority of which will rely on a deuterium and tritium fuel cycle, with a lithium-containing breeder blanket responsible for providing a sustainable tritium supply. The tritium consumption of large (DEMO-scale) devices will be in the order of 100kg per full power year, several times the current global civil tritium inventory of approximately 35kg. Whilst tritium breeding in fission reactors has supplied experimental devices (the Joint European Torus tritium inventory was limited to 90g), supporting a global fusion industry in this way is not a viable solution.

Breeder blanket technologies are immature and are yet to be tested in operational tokamak environments, where tritium breeding and extraction requirements must be met whilst maintaining structural integrity under thermal and electromagnetic loads, plasma disruption events and high neutron fluences. Therefore, it is crucial that blanket design is approached using a methodology with a proven record for success, that minimises risk and ensures thorough exploration of the associated design space.

The aim of this work is to develop blanket concept designs by utilising industry-standard systems engineering processes and methodologies. These include requirements capture, requirements validation and verification, Model-Based Systems Engineering (MBSE), Failure Modes & Effects Analysis (FMEA) and concept generation and selection methodologies.

A requirements verification plan lays out a detailed description of the verification activities that need to be executed in order to verify breeder blankets designs – including simple 1D analysis, complex transient 3D high fidelity simulations and a range of experiments and tests, forming the basis of a qualification plan.

Additionally, an analysis workflow has been developed which links an integrated set of analysis models, which provide low-fidelity concept design verification through a pre-conceptual multiphysics systems simulation, covering neutronics, thermal hydraulics, structural analysis and fuel cycle assessments.

Finally, the project has developed a methodology to provide uncertainty quantification for the key performance parameters such as tritium breeding ratio. The output is a range of down-selected blanket concepts corresponding to minimal risk and highest likelihood of meeting the sub-system (i.e. Breeder Blanket) requirements. This has allowed for the exploration of novel areas of design space, alongside the rationale which has led to the decisions.

Speaker: David Pickersgill (United Kingdom Atomic Energy Authority)

17:40 Closing of Day 1

Wednesday 3 September

08:30 Arrival

Topic I: Session 3: Chaired by Songlin Liu

Conveners: Christopher Harrington (United Kingdom Atomic Energy Authority), Helen Brooks (UKAEA), Luigi Candido (Kyoto Fusioneering Ltd), Tommi Lyytinen (VTT Technical Research Center of Finland), Yi-Hyun PARK (Korea Institute of Fusion Energy)

11 Design Exploration and Technology Development of the STEP Li2O Ceramic Breeder Blanket

The breeder blanket for the STEP Prototype Powerplant (SPP) must provide high performance breeding for a spherical tokamak without inboard breeding, materials and coolant compatible with a 600 °C outlet temperature for net power confidence, and a system deliverable on the targeted timescales of the STEP programme. Following a comprehensive assessment of all breeder, coolant, and structural material options, solid ceramic lithium oxide (Li2O) has been selected together with a Ti-modified austenitic stainless steel structural material, CO2 coolant, and beryllium-based multiplier. This combination is considered to give the highest confidence of successfully meeting the SPP requirements.
However, engineering realisation of a deployable blanket on SPP timescales still requires rapid progress in design alongside fail-fast testing and technology demonstration, with continuous iterative feedback between the two. Underpinning this must be a clear definition of requirements, constraints, and areas of uncertainty to drive robust performance development. This paper details initial scoping of the design space and the technology development needs of the chosen system.
For the chosen palette of materials, identifying a performant architecture presents initial design challenges around achieving sufficient tritium breeding performance, respecting material temperature limits through robust heat management and hydraulic design, and ensuring compliance against availability and reliability requirements. We first present and explore these constraints for the SPP blanket. For an assumed annular pin geometry, analysis revealed that peak temperatures of breeder material below 900 °C can be achieved by varying pin dimensions, but this has a consequent trade-off with structural volume content (and hence tritium breeding ratio), and total part number (and hence reliability performance). A wider exploration of potential blanket architectures is being pursued, informed by this learning, with a view to downselection of a preferred architecture.
Meanwhile, use of Li2O as a breeder material has been well documented in literature to present challenges with material degradation, most frequently citing irradiation swelling, LiOH formation, and structural material compatibility. However, these issues have a strong dependence on temperature, environment, and operational duty cycle. The issues have therefore been reviewed from first principles and revisited in the context of the SPP requirements. From this, SPP-specific design constraints and opportunities have been identified that further refine understanding of the Li2O breeder blanket design space and feed back into the design process.
Remaining uncertainties and risks have led to a set of prioritised steps for testing and technology demonstration. Preliminary screening of the extent of degradation mechanisms (and sensitivity to operationally controllable parameters) can be carried out in unirradiated environment, before more costly and time-intensive irradiation tests are required. In the longer term, scale-up towards component-level functional testing is required, aiming for proof of mechanical, thermal, and electromagnetic performance demonstration, in parallel with nuclear and tritium transport performance demonstration. A timeline for this suite of development needs will be presented to give an overall outlook for development of the Li2O blanket concept.

Speaker: Christopher Harrington (United Kingdom Atomic Energy Authority)

12 Progress and Challenges in Structural and Functional Materials Development for Breeding Blanket in Korea

The breeding blanket is one of the key components for the realization of fusion energy. It plays multiple roles, including tritium breeding for fuel self-sufficiency, heat extraction for power generation, and neutron and gamma-ray shielding for the protection of other reactor components. Reduced Activation Ferritic/Martensitic (RAFM) steel is the primary candidate for the structural material, and lithium-based ceramics are considered promising solid tritium breeder materials.
In Korea, the Advanced Reduced Activation Alloy (ARAA) has been developed by the Korea Institute of Fusion Energy (KFE) and the Korea Atomic Energy Research Institute (KAERI) since 2012. ARAA includes a small amount of zirconium to enhance impact and creep resistance. Approximately 6 tons of ARAA have been successfully fabricated on an industrial scale using Vacuum Induction Melting (VIM) and Vacuum Arc Remelting (VAR) processes. The physical, thermal, and mechanical properties of hot-rolled ARAA plates have been evaluated according to ASTM and EN ISO standards. The resulting database has been submitted to the RCC-MRx subcommittee for codification in the 2025 edition. Neutron irradiation testing and post-irradiation examination (PIE) of ARAA are currently underway using the HANARO research reactor and IMEF hot cell facilities.
For tritium breeder materials, a slurry droplet wetting method has been developed to fabricate Li₂TiO₃ and Li₄SiO₄ pebbles. An automated slurry dispensing system has also been established for mass production. The physical, thermal, mechanical, and thermomechanical properties of Li₂TiO₃ pebbles and pebble beds are being evaluated through domestic and international collaboration. Notably, neutron irradiation and tritium release tests are being conducted using the HINEG-CAS D-T neutron source and tritium handling facility at INEST in China under the Korea–China collaborative program.

Speaker: Yi-Hyun PARK (Korea Institute of Fusion Energy)

10:20 Coffee break

13 Accelerating Breeder Blanket Design with Scalable Multi-physics Modelling and Digital Engineering Workflows

Fusion breeding blankets must fulfil multiple competing objectives: maximising the extraction of both tritium and heat, minimising damaging irradiation to less resilient components (such as magnets), whilst maintaining temperatures, stresses and radiation doses below safe operational limits. To reduce costs and timescales, exploring those trade-offs efficiently is desirable, mandating an integrated modelling approach. It is necessary to incorporate several models arising from disparate domains, including neutronics, materials science, heat transfer, solid mechanics, fluid dynamics, electromagnetism, and magnetohydrodynamics.

Recognising that such a system is governed by physical regimes that are both multi-scale and non-linear implies that emergent (and potentially surprising) behaviour must be anticipated. Since transport phenomena in particular are sensitive to geometry, retaining a high degree of fidelity is also required, culminating in a large number of degrees of freedom. Therefore, it is sensible to ensure that software implementations may be easily deployed and scalable upon high-performance computing architectures.

Furthermore, the extreme conditions to which plasma-facing components are subjected (such as high neutron fluxes and thermo-mechanical gradients) are not commonly replicated within experiments, and thus data coverage of those conditions is sparse. To ensure confidence in any conclusions derived from simulation will require sensitivity analysis of stochastic and empirical model parameters, as well as the flexibility to interchange between qualitatively different theoretical approaches. Where such uncertainties are intolerable, this can lead to targeted design of experiments at prospective facilities, such as the UK Atomic Energy Authority’s future breeder mock-up test facility LIBRTI.

In recognition of these challenges, a scalable multi-physics simulation framework, suitable for modelling of breeder blankets and prospective tritium production experiments, has been developed at the UK Atomic Energy Authority. In this contribution we review its current status, commenting on the requirements of fidelity, scalability and flexibility. The MOOSE (Multi-physics Object-Oriented Simulation Environment) finite element library, with its native modules for solid mechanics and heat transfer, is taken as a starting point. This is integrated with a number of MOOSE-derived applications, including TMAP8 for tritium transport, and a coupling to OpenMC for Monte Carlo neutronics. We consider not only tritium production but also its permeation and retention in materials and explore implications for extractability.

Beyond capability for the analysis of individual design points, we explore additional software developments that facilitate a systematic approach to design exploration and qualification. Taking a Helium-Cooled Pebble Bed (HCPB) blanket concept as a test bed, we demonstrate a digital engineering pipeline to perform an optimisation over geometric parameters using active learning, as well as a sensitivity analysis to model parameters governing tritium permeation. We close with a perspective on the opportunities and challenges to improve upon this multi-physics approach to modelling and design of tritium breeding blankets.

Speaker: Helen Brooks (United Kingdom Atomic Energy Authority)

14 Advancing Neutronics Modeling of Stellarators Using Mesh-based Serpent2 Workflow

The development of next-generation stellarators is rapidly advancing, driven by recent progress in stellarator optimization that enables enhanced MHD stability and reduced turbulence [1]. Alongside public research initiatives, private companies are also pursuing pilot power plant concepts based on these designs. For a fusion reactor to operate sustainably, it must include a breeding blanket and a closed tritium cycle to ensure that tritium production exceeds its consumption. Beyond tritium breeding, the breeding blanket plays a vital role in converting the kinetic energy of fusion neutrons into heat for electricity generation and in shielding the vacuum vessel and superconducting coils from neutron damage. Its design is driven by the selection of viable materials that can withstand intense neutron irradiation and thermo-mechanical loads, while maintaining sufficient tritium breeding performance. In stellarators, additional constraints arise from the complex three-dimensional geometry—particularly the limited space between the plasma and coil systems—and the divertor configuration. These impose geometric and integration constraints that must be addressed without compromising blanket coverage, remote maintenance access, or reactor longevity.

This work presents a comprehensive overview of stellarator neutronics activities conducted using the Serpent2 [2] Monte Carlo code. The modeling workflow takes the last closed flux surface (LCFS) and coil filament current lines from the stellarator optimization as inputs, fitting the reactor layers within the space defined by the LCFS and coil system. The model also includes a layered island divertor with homogenized material composition. The geometry generation produces CAD-based STL triangle meshes that are directly compatible with Serpent’s transport routine. Tritium breeding performance is evaluated for the HELIAS 5B [3] stellarator design with different non-uniform thickness blanket configurations that balance coil shielding and breeding performance by maximizing the breeding zone volume while accommodating an additional shielding layer to peak flux regions. In addition, the impact of various divertor configurations—such as placement, area, coolants, and material composition—is assessed through a dedicated parameter study, expanding on Ref. [4] to cover the new non-uniform blanket thickness configurations. While achieving both sufficient tritium breeding and effective coil shielding is challenging with uniform-thickness layers, the tools presented in this work for generating non-uniform reactor layer thicknesses provide a promising path to meeting these requirements simultaneously.

[1] A. Goodman, P. Xanthopoulos, G. Plunk, S. Håkan, C. Nührenberg, C. Beidler, S. Henneberg, G. Robert-Clark, M. Drevlak et al., PRX Energy 3 2 (2024), 023010.
[2] J. Leppänen, V. Valtavirta, A. Rintala, R. Tuominen, EPJ - Nuclear Sciences & Technologies, 11 (2025).
[3] F. Warmer, V. Bykov, M. Drevlak, A. Häußler, U. Fischer, T. Stange, C.D. Beidler, R.C. Wolf, Fusion Engineering and Design, 123 (2017), 47-53.
[4] T. Lyytinen, A. Snicker, T. Bogaarts, F. Warmer, Fusion Engineering and Design, 216 (2025), 115000.

Speaker: Tommi Lyytinen (VTT Technical Research Center of Finland)

15 Thermal-Mechanical Analysis and Anisotropy Analysis of SiC Composites for Low-stress Scylla Breeding Blanket Modules

Speaker: Luigi Candido (Kyoto Fusioneering Ltd)

12:20 Lunch

Topic II: Session 1: Chaired by Giacomo Aiello

Conveners: Masaru Nakamichi (MiRESSO Co. Ltd.), Milan Zmitko (Fusion for Energy (F4E)), Pierre Lamagnère (CEA), Takanori Hirose (National Institutes for Quantum Science and Technology), Wanjing Wang (Wanjing Wang, Associate Professor, ASIPP)

14:00 Introduction to Topic 2

16 European Test Blanket Modules: An Overview of Fabrication Technologies Development and Their Feasibility Assessment

Within the framework of European fusion strategy, two reference tritium Breeder Blankets concepts are developed to be tested in ITER as Test Blanket Modules (TBMs): Water-Cooled Lithium-Lead (WCLL) which uses liquid Pb-16Li as a breeder and neutron multiplier, and Helium-Cooled Ceramic Pebble (HCCP) with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier materials. Both concepts use as a structural material Reduced Activation Ferritic Martensitic steel, EUROFER97 (X10CrWVTa9-1) [1]. Pressurized water (15.5 MPa, 295-328ºC for WCLL) and pressurized helium (8 MPa, 300-500ºC for HCCP) are used for heat extraction.
The TBM structure is constituted of a Box (made of two Side Caps (SC) and First Wall (FW)), stiffened by horizontal and vertical Stiffening Plates (SP) and closed on its back, in the manifold area, with several Back Plates (BP) of different thicknesses with passing through elements. Inside the Box, cooled elements, Double Wall Tubes (DWT) for WCLL concept and Cooling Plates (CP) for HCCP concept, are assembled into breeder units delimited by Stiffening Plates. All HCCP structural subcomponents are internally cooled with He circulating in meandering square section channels. Similar concept is used for the water-cooled First Wall of the WCLL.
This paper briefly describes the strategy adopted for the development of the TBM-related fabrication technologies, and an approach used for manufacturing and delivery of the TBMs on ITER, considering particularities of EUROFER97 structural material and relevant regulatory aspects (TBM is classified as a Nuclear Pressure Equipment) [2]. The main characteristics of the EUROFER97 structural material are given in this context [3].
Fabrication technologies developed to (i) manufacture TBM subcomponents (FW, SC, SPs, CPs), (ii) assemble TBM Box, (iii) assemble TBM Box structure, and (iv) assemble TBM back manifold, are overviewed and discussed [4-6]. The applied welding technologies are based on fusion welding (laser beam, GTAW) and diffusion bonding (HIP) considering specificities of EUROFER97 steel. Preliminary Welding Procedure Specifications (pWPS) are developed following requirements of professional standards and RCC-MRx construction code [7].
Moreover, the paper presents the main outcomes related to the phase of consolidation of welding processes and related technologies for manufacturing of EUROFER97 structures and components. It discusses (i) EUROFER97 weldability demonstration for various welding techniques used for the TBM manufacturing (like GTAW/TIG, EBW, LBW and HIP), (ii) filler material acceptance studies, (iii) assessment of the effect of multiple Post-Weld Heat Treatment (PWHT) on mechanical properties of weld joints, (iv) dissimilar welding between EUROFER97 and 316L(N)-IG stainless steel, and (v) Double-Wall Tubes (DWT) manufacturing and DWT/BP weld joint development and characterization.
The conclusions of the paper discuss the main encountered issues and challenges related to the TBM manufacturing feasibility and relevant Return-of-eXperience (RoX).

Speaker: Milan Zmitko (Fusion for Energy (F4E))

17 Characterization Program of EUROFER97 RAFM Steel and Implementation Plan in the RCC-MRX Code for the Design and Manufacturing of ITER TBM

A set of different Test Blanket Module (TBM) concepts will be installed on ITER to validate the design and operation in nuclear fusion environment of Breeding Blanket technologies for fusion facilities. AFCEN code RCC-MRx [1] has been selected for the design and the manufacturing of the European TBMs. Reference Procurement Specification and some material properties are already included in the “probationary rules” tome of the RCC-MRx code for the reduced activation ferritic martensitic (RAFM) steel EUROFER97 [2] chosen for the structural material of European TBMs. In order to fill the remaining gaps needed to complete the implementation of EUROFER97 in the RCC-MRx code [3], an extensive experimental programme is being conducted by the EUROfusion Consortium. On this basis, the codification work in RCC-MRx has recently restarted in a collaboration with Fusion for Energy (F4E), Framatome and CEA.

Qualification programme for EUROFER97
Seven laboratories collaborating within EUROfusion are involved in the qualification programme [4]. Progress of the 2020-2025 programme for the qualification of base metal, including the validation of design rules and evaluation of the effect of irradiation will be presented with a focus on ratcheting, creep-fatigue interaction and immediate plastic flow localization. The 2025-2027 programme for the qualification of tungsten inert gas (TIG) welded joints is starting with tests planned before and after neutron irradiation.

Implementation in the RCC-MRx code
Codification activities driven by F4E are restarting through a five-year framework contract by a consortium FRAMATOME/CEA. The objectives of the contract starting from 2025 are to:
— Analyse the data from the EUROfusion experimental programme in view of preparing the next Modification Request for integration of EUROFER97 in the Tome 1 (Design) of the RCC-MRx.
— Update and complete Reference Procurement Specification in the Tome 2 with experience gained from the manufacturing of the 4th batch of EUROFER delivered by Saarschmiede with certificates EN 10204 3.1.
— Consolidate the codification for the manufacturing and control of TBMs, including welding (Tomes 3 to 5) using the feedback from the development activities for the fabrication and assembly processes.

The EUROfusion experimental programme, as well as the progress on procurement and fabrication development by F4E, will allow completing the supporting Material File requested to justify the design and manufacturing of the TBMs according to the requirements of the RCC-MRx code.
REFERENCES
[1] AFCEN, RCC-MRx, 2022.
[2] A.A TAVASSOLI et al., Journal of Nuclear Materials 455 (2014) p. 269-276.
[3] Guide for introducing a new material in the RCC-MRx. Technical Publication form AFCEN, 2017.
[4] D. TERENTYEV et al., 29th IAEA Fusion Energy Conference 16-21 Oct 2023 London, UK (2023).

Speaker: Pierre Lamagnère (CEA CADARACHE)

18 Beryllium Resource and it’s Stably Securing

In the Fusion Energy, large quantities of beryllium are essential. Tritium as fuel is self-produced in the blanket covering the plasma during operation. Beryllium is essential for the blanket as it multiplies neutrons and produces more tritium. Fusion reactors require more than 350 metric tons of beryllium based on the blanket design. However, there are big issues for beryllium procurement. One is overpriced, other is insufficient production in the world.
MiRESSO has the novel process with cost reduction. MiRESSO has established a low-energy dissolution with low temperature at 300˚C. It is the combination process between a base fusion method for dissolution of insoluble substances and a microwave heating. In this first process, Beryl as insoluble composition is changed to soluble composition. Then, in the next process, Beryllium completely dissolve by acid solution at room temperature under normal pressure.
Business outlines of MiRESSO are “Production and sales of Beryllium” and “Licensing & consultation for energy saving and CO2 emission reduction of refining and recycling process with high temperatures”. We aim to do a contribution to social and economic security by stably securing mineral resources. MiRESSO came from Mineral Refining and Recycling System Society and established last year.

Speaker: Masaru Nakamichi (MiRESSO Co. Ltd.)

15:50 Coffee break

19 R&D Status in Manufacturing and Assembly of Tritium Breeding Blanket Component

Tritium breeding blanket is the core component in the future fusion power plants, which is responsible for tritium breeding, neutron shielding and heat extraction. In order to realize these functions, the blanket component usually is designed as a box-structure modular made of RAFM steel plates, into which the tritium breeding materials and neutron multiplying materials were filled. To exhaust the nuclear heat in the blanket, a serial of square cooling channels was installed in the RAFM steel plates. At the same time, in order to improve the thermal and particle impact resistance on blanket components, tungsten armor is selected as the plasma-facing material for the first-wall sub-component. Therefore, the manufacturing and the assembly of tritium breeding blanket components involves the forming of internal cooling channels in RAFM steel plates, the joints of tungsten /RAFM steel dissimilar material and the welding of RAFM steel and RAFM steel. On the other hand, preparation techniques of reliable tritium permeation barrier (TPB) on RAFM materials need to be explored to prevent tritium loss in the cooling water. To solve these issues, ASIPP sets up a special engineering research team to carry out the development and research of the manufacturing technology for tritium breeding blanket module.
In this report, we will show some recent research progress in the manufacturing and assembly of blanket module. Firstly, we will outline the manufacturing issues for the sub-components in the CFETR-WCCB module, such as the First Wall (FW), the double-wall tubes (DWT) and the manifold [1, 2]. Then the R&D activities for the machining and fabrication of these sub-components have been carried out. In the following, the assembly welding process of the WCCB modules was proposed, and the technology issues for the full-size module were solved. Finally, the first full-size WCCB module was successfully manufactured in ASIPP [1, 2]. In this processing, we set up a number of large-scale machining, welding device and testing platforms. At the same time, some new bonding technology of Fe-Cr-Al and RAFM steel has been developed to prepare the reliable TPB in the cooling channels [3, 4]. And the first FW mock-up with TPB in the cooling channels has been fabricated. In addition, the 3D printing technology has also been applied in the manufacturing of blanket component, and some result will be provided in this presentation.

Speaker: Wanjing Wang (Wanjing Wang, Associate Professor, ASIPP)

20 Overview of Japanese Water-Cooled Ceramic Breeder Test Blanket Module Manufacturing and Assembly Technologies

The Water-Cooled Ceramic Breeder (WCCB) blanket is considered a promising and reduced-risk technology for fusion reactors, owing to its extensive operational experience in pressurized water fission reactors and the well-established understanding of material behavior under neutron irradiation [1,2]. This blanket concept offers a viable pathway for the early realization of energy conversion and tritium self-sufficiency, both of which are essential for the development of sustainable fusion energy systems.
As part of the ITER Test Blanket Module (TBM) project, which aims to experimentally validate breeding blanket technologies for future demonstration reactors, Japan is preparing to test a WCCB-TBM. The Japanese module employs high-temperature, high-pressure water at 15.5 MPa/280°C as the coolant, and incorporates pebble-type materials for tritium breeding (Li₂TiO₃) and neutron multiplication (Be) [3,4]. The structural material is F82H, a reduced-activation ferritic/martensitic steel developed from 9Cr heat-resistant steel, selected for its favorable mechanical properties and radiation resistance [5].
To ensure both safety and thermal performance under fusion-relevant conditions, the Japanese WCCB-TBM adopts a cylindrical structural design with its radial axis serving as the major axis. This configuration enables stable containment of high-pressure coolant even in the event of internal leakage and facilitates rapid heat dissipation to the vacuum vessel or surrounding structures in case of cooling system failure [6]. The design reflects a careful balance between structural integrity, manufacturability, and functional performance [7-9].
This work provides a comprehensive overview of the manufacturing and assembly technologies developed for the WCCB-TBM. Key manufacturing processes include machining of F82H components, assembly, joining and pebble packing. [10]. From the perspective of pressure resistance and vacuum compatibility, welded joints are required in the blanket structure. Since F82H is a tempered martensitic steel, post-weld heat treatment is necessary to restore mechanical strength. However, if this heat treatment is performed after the functional materials have been filled, it may promote the formation of reaction layers at the contact interfaces between F82H and the functional materials. Moreover, repeated heat treatment can also lead to degradation of the weld metal strength. Therefore, careful consideration of the assembly sequence is essential.
REFERENCES
[1] KAWAMURA, Y, et al., Fusion Engineering and Design, 201 (2024) 114260.
[2] OCHIAI, K., et al., Fusion Engineering and Design 89, Issues 7-8 (2014) 1464-1458.
[3] HOSHINO, T., et al., Fusion Engineering and Design 109-111, Part B (2016) 1114-1118.
[4] NAKAMICHI, M., et al., Fusion Engineering and Design 88, Issues 6-8 (2013) 611-615.
[5] NOZAWA, T., et al., Nuclear Fusion, 61, (2021) 116054.
[6] TANIGAWA, H., et al., Fusion Engineering and Design 136 (2019) 1221-1225.
[7] GUAN, W., et al., Fusion Engineering and Design 190 (2023) 113637.
[8] GUAN, W. et al., Fusion Engineering and Design, 200 (2024) 114202.
[9] HIROSE, T. et al., Fusion Engineering and Design, 85, Issues 7-9 (2010) 1426-1429.
[10] HIROSE, T. et al., Fusion Engineering and Design, 200 (2024) 114227.

Speaker: Takanori Hirose (National Institutes for Quantum Science and Technology)

Topics I, II, III Posters: Poster session

All posters from Topics I,II,III will be presented within the Poster Session

18:50 Reception

20:00 End of Day 2

Thursday 4 September

08:30 Arrival

Topic III: Session 1: Chaired by Hiroyasu Tanigawa

Conveners: Angel Ibarra (CIEMAT), Giacomo Aiello (EUROfusion), Ivo Moscato (EUROFUSION Consortium), Kecheng Jiang (Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences), Paul Humrickhouse

21 The Role of Different Facilities for the Nuclear Qualification of the BB

This work evaluates the roles of ITER, DONES, and the recently proposed European Volumetric Neutron Source (VNS) in achieving the nuclear qualification of the Breeding Blanket (BB) within the European fusion roadmap. Using a Technology Readiness Level (TRL) framework, it identifies how experimental testing progresses from basic material validation (TRL 4) to full system demonstration (TRL 8). The study shows that while DONES provides valuable early-stage data and ITER supports mid-level integration tests, only a VNS enables complete qualification at TRL 8, involving full-scale segments under realistic fusion conditions. The analysis maps BB qualification needs to facility capabilities, identifies critical gaps, and proposes strategic use of each facility. It concludes that VNS is essential to reduce risks, delays, and uncertainties in DEMO deployment, supporting a reliable path to commercial fusion energy. This conclusion aligns with similar strategies proposed by international actors such as the U.S. FNSF and China’s CFETR.

Speaker: Giacomo Aiello (EUROfusion)

22 Testing and Nuclear Qualification Strategy for CFETR Breeding Blanket

Breeding Blanket is in charge of breeding tritium for tritium self-sufficiency, shielding neutron for environment protection and extracting fusion energy for electricity generation. There are three candidate blankets for Chinese Fusion Engineering and Test Reactor (CFETR), including the Helium Cooled Ceramic Breeder (HCCB) Blanket, Water Cooled Ceramic Breeder (WCCB) blanket and supercritical CO2 cOoled Lithium-lead (COOL) blanket. The operating conditions and functional requirements for blanket are critically harsh, and this is resulted from the multiple effects of nuclear, electromagnetic, thermal hydraulic, thermal-mechanical as well as the tritium safety issues. Installation of a breeding blanket in CFETR without prior fusion testing is found to result in high risks of not attaining the required tritium self-sufficiency, blanket system reliability and an adequate device availability. Therefore, in the support of Comprehensive Research Facility for Fusion Technology (CRAFT), the testing and nuclear qualification strategy for CFETR blankets will be performed, as the followings: (1) The water / S-CO2 loop with high temperature and pressure will be used for thermal hydraulic test on the blanket structural components, i.e. High Heat Flux (HHF) test on the FW with one-side heating, Critical Heat Flux (CHF) occurrence, flow distribution into different components and channels, as well as the flow instability for parallel channels; (2) The PbLi loop can be used to study the MHD effects, including the phase diagram/turbulent transition mechanism under different Re/Ha, MHD flow in complex geometry channel and multi-channels under electromagnetic coupling effects. The more importantly, it is can be further used to test the ability of FCIs minimizing the pressure drop under magnetic field, and quantify the performance of material corrosion resistance; (3) Using the pebble beds experimental platform, we can test the flow / heat transfer performance between the purge gas and pebble beds, characterizing the thermal expansion stress using the thermal-mechanical facility, mixing and sieving the pebbles, and analyzing the packing structure using CT, as well as carry out the packing structure stability using the vibration facility; (4) Using the Oxide Dispersion-Strengthened (ODS) alloy ferritic steel, the full-sized blanket module will be manufactured and hydraulic tested through the 3D printing technology, and this aims to reduce the welding joints for high thermal-mechanical strength; (5) The nuclear test for tritium generation for the scaled mockup of WCCB has been finished, and the test for COOL blanket is under going, in which the Tritium Production Rate (TPR) will be measured using the DT neutron source, combined with the 6Li glass on-line measurement lithium glass detector and offline liquid scintillator. This aims to understand the uncertainty degrees deviation from the target of tritium self-sufficiency, i.e. the nuclear reaction cross-section, analysis codes and the spacial distribution of neutron source. Therefore, the above various testing and qualification will ensure the confidence of attaining the required tritium self-sufficiency and system reliability before the installation of blanket into the CFETR.

Speaker: Kecheng Jiang (Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences)

10:20 Coffee break

23 IFMIF-DONES: Experimental strategy for fusion materials qualification

IFMIF-DONES (International Fusion Materials Irradiation Facility – DEMO Oriented Neutron Source) has been identified as a strategic facility in the European roadmap toward the achievement of fusion energy. Currently under construction in Granada (Spain) and expected to be operational in 2034, it will be the only installation worldwide capable of producing a high-intensity, fusion-relevant neutron flux based on deuteron–lithium (Li(d,n)) reactions. This flux, expected to reach densities of up to 10¹⁵ n/(cm²·s), will be delivered in the central Test Cell through a modular configuration designed to accommodate multiple irradiation environments. The resulting neutron spectrum will closely reproduce that of a fusion reactor, enabling the accelerated testing of materials under well-controlled thermal and mechanical conditions.
The main objective of IFMIF-DONES is to enable the nuclear qualification of structural and functional materials for DEMO and future fusion reactors. The facility will deliver up to 20–30 dpa in 300 cm³ within 2.5 full-power years (fpy) and 50 dpa in 100 cm³ in 3 fpy. In its modular configuration, the primary irradiation system will be the High-Flux Test Module (HFTM), optimized for small specimen testing (SSTT) and initially dedicated to Eurofer97 and other RAFM steels, providing fundamental data on irradiation-induced hardening, embrittlement, microstructural evolution, and helium effects. IFMIF-DONES include other modules under development such as the Creep Fatigue Test Module (CFTM), addressing fatigue life reduction under neutron irradiation, the Tritium Release Test Module (TRTM), to quantify tritium release behavior from ceramic breeder materials, and other flexible irradiation modules for specialized experiments.
IFMIF-DONES is also planned to support complementary experimental applications in fields such as medical isotope production, gamma spectroscopy, time-of-flight measurements, and neutron-based industrial processes.
IFMIF-DONES will play a key role in the European fusion roadmap by providing the data needed to evolve from material development to operational components in future fusion reactors.

Speaker: Angel Ibarra (IFMIF-DONES)

24 Progress in the Concept Development of the VNS - A beam-driven Tokamak for Component Testing

The volumetric neutron source (VNS) is a compact beam-driven tokamak with D-T plasma to generate a high neutron flux that will allow the testing and qualification of fusion nuclear components, in particular the breeding blanket. Recently, EUROfusion concluded a feasibility study that confirmed the principal feasibility of the machine and the plant for construction and operation. Also, aspects were identified that require further development and assessment, which have been key subjects of the conceptual design phase that has been on-going since.
This article summarizes the progress made in the design of VNS, including (i) the rationale for the minor modifications of major radius and aspect ratio, (ii) the configuration and performance of the plasma equilibrium coils, (iii) the design of the in-vessel components including their remote handling concepts, (iv) the concepts of the neutral beam and electron cyclotron heating systems, (v) integration concepts of the main plant systems in particular those of the tritium fuel cycle. In addition, the article provides an outlook on the expected operation plan and summarizes the potential of VNS for testing and qualifying the components and technologies required for a fusion power plant.

Speaker: Ivo Moscato (EUROFUSION Consortium)

25 Overview of Some Planned Blanket Test Facilities in the US

Speaker: Paul Humrickhouse

12:30 Lunch

Topic III: Session 2: Chaired by Salvatore D'Amico

Conveners: Dmitry Terentyev, Mark Gilbert (CCFE), Mu-Young Ahn (Korea Institute of Fusion Energy)

26 The IFMIF-Dones Test Blanket Units

Breeding blanket designs considered for DEMO include not only materials but also technologies, whose behaviour under fusion-like conditions has not yet been tested. Therefore, it is urgent to evaluate these blankets in relevant environments, with emphasis on significant radiation levels. Looking for solutions to qualify and validate the breeding blankets, IFMIF-DONES has launched a new experimental program on Test Blanket Units (TBU). The primary objective is to contribute to the BB testing in an irradiation environment similar to that expected in a fusion reactor, by performing multi-physics experiments, and highlighting the capabilities of IFMIF-DONES to qualify tritium technologies in its medium-flux area.
A preliminary exercise has already been performed with the Helium-Cooled Pebble Bed (HCPB) and the Water-Cooled Lead Lithium (WCLL) blankets, demonstrating that the effective irradiation volume in IFMIF-DONES is sufficient for relevant tritium experiments. Additionally, the TBU can help demonstrate effective temperature control of the blanket or test the bonding quality between different materials or components under a high neutron flux.
IFMIF-DONES already comprises dedicated spaces to host the auxiliary systems necessary for a proper operation of the TBU, including those for tritium handling, as well as other supplies and services (e.g. cooling loops, power supply…). The auxiliary systems will, in turn, monitor the purity of the tritium carrier (gas or liquid) and will be compatible with the additional needs in terms of detritiation and tritium storage that the IFMIF-DONES plant will provide. The current baseline of the main irradiation area, the Test Cell, already considers supplies and services (via the PCPs, Piping and Cabling Plugs) to the modules and TBUs that will be positioned behind the lithium target. Some of these PCPs are already planned to extend to the rear wall of the TC Liner.
In summary, together with the TBM-Program and a possible validation of blankets in a future VNS, it is expected that IFMIF-DONES can help increase the TRL of this important component.

Speaker: David Rapisarda (CIEMAT)

27 Irradiation of Be- and Li-Based Materials for Application in ITER TBM

A set of different Test Blanket Module (TBM) concepts will be installed on ITER to validate the design and operation in nuclear fusion environment of Breeding Blanket technologies for fusion facilities. The European solid breeder blanket concept HCPB to be tested at ITER uses advanced ceramic breeder (ACB) material containing lithium (Li) in the form of pebbles and Beryllium (Be) material in the form of pebbles [1].
Beryllium is required for neutron economy. In addition to pure Be, the intermetallic compounds such as TiBe12 and CrBe12 beryllides were proposed as alternative solution thanks to their improved properties, including lower swelling, superior mechanical strength, and better suitability for industrial-scale production. The first irradiation study of Be and TiBe12 back in 2000-2005, is reviewed in [2] and it highlights superiority of the titanium beryllide over pure Be. Recently, a new batch of Be, TiBe12 and CrBe12 has been produced and its qualification including response to the neutron irradiation is necessary.
For the fuel supply, the biphasic Li4SiO4/Li2TiO3 ceramics are proposed as reference tritium breeder ceramics for the European solid breeder blankets for ITER and DEMO [3]. The advantage of these composites is resistance of grain growth against long-term annealing and maturity of the high-throughput production beyond lab scale. Just as in case of newly produced beryllides, the investigation of the response of the advanced Li ceramic pebbles to the neutron irradiation and prove of the breeding capability is required.
This presentation reviews the latest irradiation experimental campaign which has been set at Belgian Nuclear Research Centre as one of the elements in the qualification process of functional materials for ITER TBM. The talk will cover the main elements of design space (scope-time-budget), safety aspects (tritium, chemical compatibility, build-up of gas pressure) to deploy the experiment and during the experiment, cross-border nuclear transportation aspects and finally waste disposal considerations.

References:
[1] F.A. Hernandez et al., Fusion Energy Technology R&D priorities (2025) 225-234.
[2] V. Chakin et al., Nuclear Materials and Energy 42 (2025) 101910.
[3] O. Leys et al., Fusion Engineering and Design 164 (2021) 112171.

Speaker: Dmitry Terentyev

28 The science objectives and basis of the LIBRTI tritium breeding test facility

Fuel (tritium) self-sufficiency is a critical requirement for the successful realisation of fusion as a viable energy source. More urgently, even the next generation of fusion experiments and prototype plants may be unable to rely on non-fusion sources of tritium to supply start-up or top-up inventories of tritium, in other words, they are likely to need to be self-sufficient in tritium almost as soon as they begin operating and start burning tritium. Primarily for this reason, UK Atomic Energy Authority has been awarded ~£200M to design and build, by 2028, a facility to provide a test platform for tritium breeding technologies to enable the research needed to derisk the solutions for tritium breeding for UK’s own prototype power plant, STEP, as well as for global fusion development. The facility will be based around a compact neutron source, initially delivering of the order of 10$^{13}$ n/s from DT reactions. Tritium breeding mock-up experiments at the m$^3$ scale are being developed in partnership with fusion industry, while a multiphysics digital modelling platform is being constructed in parallel – a digital platform that will be used to guide experimental design and be subsequently validated in its predictions of tritium production and recovery by the experimental measurements on LIBRTI.
This paper will describe the science basis for LIBRTI, explaining the motivation behind the programme and detailing a preliminary concept for one possible experiment, which has been used to validate the measurability of tritium production from LIBRTI experiments. These initial model tests demonstrate that not only will LIBRTI experiments produce measurable amounts of tritium, but, further, that experiments will be able to access time-resolved tritium recovery rates without being contaminated by the adjacent, tritium-based neutron source. This is an important validation of the engineering design choices for the facility, which will begin construction in 2025.

Speaker: Mark Gilbert (CCFE)

29 Test Strategy and Infrastructure for Breeding Blanket Development at KFE

The development of breeding blankets is critical for the realization of fusion energy, as they are essential in fuel production and energy generation in fusion reactors. The pre-conceptual design for the K-DEMO blanket has commenced, with the HCCP (Helium-Cooled Ceramic Pebble) blanket concept adopted as the reference design following the KO-EU HCCP TBM project, while other potential design options are being explored. To support and validate these designs, studies have been conducted to derive the strategy and infrastructure necessary for breeding blanket development in Korea.

As part of such efforts, Korea Institute of Fusion Energy (KFE) performed a pre-conceptual study of the Korea Fusion Engineering Advanced Test Complex (KFEAT) which encompasses three primary facilities: the Integrated Breeding Test Facility, the Blanket System Test Facility, and the Fuel Cycle Pilot Facility. At the core of this proposal, the Integrated Breeding Test Facility is designed to perform component-level testing of breeding blankets for validation of overall performance under DEMO-relevant irradiation time and scenarios. It is based on a 40 MeV deuteron accelerator-driven system operating at up to 10 mA, capable of generating fusion-like neutrons. This enables neutron testing of blanket breeding units, potentially at one-to-one scale.

In the meantime, with the introduction of the “Strategy for Accelerating Fusion Energy Realization” in 2024, the development of key fusion technologies, including breeding blankets, is expected to gain further momentum. In alignment with this strategy, KFE has proposed a compact pilot device that will demonstrate steady-state operation and fusion-relevant performance in an industrially scalable form. Although the design requirements and key parameters of the device are still under discussion, it is envisaged that breeding blankets will be operated from early stage to supply tritium necessary for sustaining the fusion reaction.

While the development of the compact pilot device and supporting strategies is underway, the establishment of appropriate infrastructure for the qualification of breeding blankets remains a critical challenge. In this context, discussions are ongoing regarding how the compact pilot device and facilities such as KFEAT can be strategically aligned and utilized in a complementary manner, including the possibility of using the device as a volumetric neutron source. The outcome of these discussions is expected to shape Korea’s long-term fusion roadmap and guide the development of breeding blankets.

Speaker: Mu-Young Ahn (Korea Institute of Fusion Energy)

16:00 Coffee break

General Discussion: Topics I, II, III: Chaired by Luciano Giancarli

18:00 Closing remarks

Friday 5 September

Closed session: Programme committee only