22nd International Workshop on the Ceramic Breeder Blanket Interactions (CBBI-22)

Oct 6, 2025, 9:10 AM → Oct 8, 2025, 6:00 PM Asia/Tokyo

Kihada Hall (Obaku Plaza) (Kyoto University Uji campus)

Keisuke Mukai

Welcome to the 22nd International Workshop on the Ceramic Breeder Blanket Interactions (CBBI-22), the biannual event that brings together the community of ceramic breeder scientists and technologists. 

It is the forum for the researchers, engineers and technologists, involved in the development of the fusion ceramic breeding blanket concept, to exchange the latest progress in the design, fabrication, modeling, materials, isotope separation, and tritium breeding/extraction.

Organized under the auspices of the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and in conjunction with the International Conference on Fusion Materials (ICFRM-22) in Shizuoka, the CBBI-22 will be held from 6th to 8th October 2025 (three days), in the beautiful and historical city of Kyoto, Japan. 

Your oral presentations and topical discussions will be essential contributions to improve the ceramic breeder blanket as the system for tritium fuel production and energy extraction in a fusion energy reactor. The CBBI-22 will be held jointly with The 17th International Workshop on Beryllium Technology, and a joint session is also scheduled to take place.

This workshop should be the stage to show updates of TBMs for ITER and a debate forum for its exploitation towards DEMO and Fusion Pilot Plant (FPP) by the private sector.

Participation in this workshop is free of charge.

 

   First announcement   

Call for Abstract

Deadline for abstract submission: 28th March 2025

Abstract template: Word

 

Registration

Please register from here. All participants, including non-presenters, are kindly requested to complete registration. The presenters are requested to complete the registration by 25th July. 

A technical tour at Heliotron J plasma facility at Kyoto University Uji campus will be held on 8th October from 14:00 (about 60-90 min). If you wish to attend the technical tour please complete the tour registration by 23rd July 2025 using the link below (Note that you cannot register it after the deadline)

 

Key dates

Deadline for Abstract due ：28th March 2025　　11th April 2025 (extended)

Notification of acceptance ：6th June 2025

Registration for oral presentators: 25th July

Registration without oral presentators: 30th September 

Final Program: 31st August 2025
Workshop ：6-8th October 2025
 

Topics of the Technical Program 

1. Solid breeder blanket design for ITER-TBM, DEMO, and FPP
2. Ceramic breeder material development, production and qualification
3. Simulations and modeling for materials and pebble beds 
4. Hydrogen isotope behavior and isotope separation technology
5. Irradiation testing 
6. Topical discussions on selective key issues

 

Local Organizing Committee 

Keisuke Mukai (Workshop Chair)	NIFS
Juro Yagi	Kyoto University
Jae-Hwan Kim	QST
Takumi Chikada	Shizuoka University

Program Advisory Board 

M. Abdou	UCLA
M-Y. Ahn	KFE
L. V. Boccaccini	KIT
P. Chaudhuri	IPR
X. Chen	CAEP
K. Feng	SWIP
A. Ibarra	CIEMAT & IFMIF-DONES España
Y. Kawamura	QST
R. Knitter	KIT
T. Terai	IAE

 

Acknowledgement 

This workshop is co-organized and supported by the Joint Usage/Research Program on Zero-Emission Energy Research, Institute of Advanced Energy, Kyoto University (ZE2025D-04).

CBBI-22_AbstractTemplate.docx

First-announcement.pdf

Mon, October 6

9:10 AM Welcome

CBBI/BeWS joint session 1

1 [BeWS] Modeling and experimental validation of irradiation defects in HCP materials

Pavel VLADIMIROV

Speaker: Pavel VLADIMIROV

2 [CBBI] Current Plans for the HCCP TBM testing in ITER and Brief Overview of R&D Activities on the European Ceramic Breeder

Milan Zmitko1, Italo Ricapito2, Regina Knitter3, Julia Leys3, Salvatore D’Amico4
1 Fusion for Energy (F4E), c/ Josep Pla 2, Barcelona, Spain
2 Fusion for Energy (F4E), Route de Vinon-sur-Verdon, 13115 St Paul lez Durance, France
3 Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM), Karlsruhe, Germany
4 EUROfusion - Programme Management Unit, Boltzmannstrasse 2, Garching, Germany

One of the reference tritium Breeder Blanket concepts developed in the Europe that will be tested in ITER machine under the form of Test Blanket Module (TBM) is the Helium-Cooled Ceramic Pebble (HCCP) TBM concept in which lithiated ceramic pebbles are used as a tritium breeder and beryllium/beryllides as a neutron multiplier material. This concept, which is being jointly developed in collaboration with ITER Korea under a Partnership agreement, uses EUROFER97 reduced activation ferritic-martensitic (RAFM) steel as a structural material and pressurized helium for heat extraction (8 MPa, 300-500ºC).
The paper gives a brief general description of the HCCP TBM design and the main design requirements, including the requirements for the ceramic breeder material. The HCCP TBM development and qualification plan and ITER testing objectives with identification of the main milestones will be presented, taking into account new ITER baseline and derived from it a TBS Research Plan elaborated considering new testing conditions.
The main part of the paper will be devoted to the presentation of the ceramic breeder material development strategy, qualification plan and overview of the R&D activities. The achieved results on the ceramic breeder (CB) pebbles production (KALOS process), the CB pebbles and pebble beds characterization, and performance under TBM/DEMO relevant conditions, including the performance under neutron irradiation, and thermo-mechanical performance will be briefly overviewed and a new neutron irradiation experiment, foreseen for the functional materials (i.e. for the ceramic breeder and beryllium materials), will be introduced. A special attention will be focused on the on-going R&D activities: (i) development of the CB/Be pebbles filling procedure, and (ii) modelling of the CB pebbles fragmentation and a possible dust formation with simulation studies on its transport and deposition.

Speaker: Milan Zmitko

10:20 AM Coffee break

CBBI/BeWS joint session 2

3 [CBBI] ORNL Status and Progress in Research on Ceramic Breeder Blanket Materials

Xiao-Ying Yu*, Yutai Katoh, Takaaki Koyanagi, German Samolyuk, Osetskiy, Yuri, and Weicheng Zhong
Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA

The fusion energy sciences program at the Oak Ridge National Laboratory (ORNL) contributes in a wide range of technical areas to support fusion reactor development. In particular, topics related to the fusion blanket and fuel cycle are among the major thrusts in materials science and technology development at ORNL. Although several specific tritium breeding concepts and their materials needs have been explored, uncertainties remain on the requirements and solutions for the materials needed for the blanket components. Multiple engineering challenges are imposed by the harsh operating environment of neutron exposure, high temperature and stresses, and evolving material property changes due to the irradiation environment. We will present recent updates and progress on our research activities of specific interest for the ceramic breeder blanket concepts. The talk will cover several research topics: 1) neutron irradiation effects in lithium ceramic breeding materials, 2) silicon carbide for electrical insulation, thermal insulation and tritium permeation control, 3) tritium surrogate diffusion behavior under fusion-relevant conditions, 4) atomistic modeling to improve understanding of solid breeders, and 5) advanced manufacturing of ceramics of high potential for adoption in fusion reactors. Currently active work includes a) atomistic modeling of structural stability and tritium trapping/detrapping in irradiated Li2TiO3 and Li2ZrO3, b) hydrogen isotope permeation and trapping in candidate membrane, and c) upcoming opportunities of the FIRE center research activities that will focus on tritium breeding, accountancy, and breeder material development and validation in the US. Future expansion will include irradiation experiments to examine the stability and properties of lithium ceramics using the established capabilities of the HFIR reactor and our suite of PIE testing and characterization equipment.

Speaker: Xiao-Ying Yu

4 [BeWS]

Jaehwan KIM

Speaker: Jaehwan KIM

5 [CBBI] Overview of Breeding Blanket Technology Development in Korea

Mu-Young Ahn1*, Seungyon Cho1, Hyoseong Gwon1, NamIl Her1, Hyung Gon Jin2,
Chang-Shuk Kim1, DoHyun Kim1, Myungho Kim1, Sungyu Kim1, Suk-Kwon Kim2,
Woong Chae Kim1, Dong Won Lee2, Yonghee Lee1, Youngmin Lee1, Sungbo Moon1,
Soon Chang Park1, Yi-Hyun Park1, Chang Wook Shin2, Seok-Kwon Son1

1 Korea Institute of Fusion Energy (KFE), Daejeon, Republic of Korea
2 Korea Atomic Energy Research Institute (KAERI), Daejeon, Republic of Korea

The development of breeding blanket is critical for the realization of fusion energy, as it plays a vital role in fuel production and energy generation in fusion reactors. The pre-conceptual design for the K-DEMO blanket has been initiated, with the HCCP blanket concept chosen as a reference design following the KO-EU HCCP TBM project, while other potential options continue to be explored. To efficiently support and validate these designs, a conceptual study has been conducted to derive the strategy and infrastructure necessary for breeding blanket development.

In the meantime, with the "Strategy for Accelerating Fusion Energy Realization" approved in Korea in 2024, it is expected to further accelerate the development of key fusion technologies, including breeding blankets. While the strategy and infrastructure for blanket development will need to be realigned in accordance with this new strategy, basic R&D activities for breeding blankets will continue. These efforts include the development of tools, modeling and data for design and safety, manufacturing technologies, tritium extraction and cooling technologies, and materials and their database.

This paper addresses the breeding blanket development strategy and provide an overview of the current status of technology development in Korea, highlighting ongoing R&D activities and key advancements in breeding blanket technologies.

Keywords: blanket, tritium breeding, DEMO, TBM

Speaker: Mu-Young Ahn

12:20 PM Lunch break

Tritium transport

Convener: Chase Taylor

6 Preliminary Study on Tritium Release Properties of Li2TiO3 Pebbles Manufactured by PIM Process

Yi-Hyun Park1, Young Ah Park1, Mu-Young Ahn1, Haixia Wang2
1KFE, Korea, 2INEST, China

The solid type of tritium breeder is used as a pebble form due to packing phenomena, stress concentration, thermal conductance, purge gas flow, and so on. The important properties of breeder pebbles are not only physical properties but also tritium release property. This study aims at a preliminary investigation on tritium release properties of lithium metatitanate (Li2TiO3) pebbles manufactured by Powder Injection Molding (PIM) process.
Li2TiO3 pebbles with 3.4 mm in diameter were prepared by PIM process which was first try to pebble manufacturing by Korea Institute of Fusion Energy (KFE). Although Li2TiO3 pebbles with around 1 mm in diameter are foreseen to be used in breeding blanket, those with 3.4 mm in diameter were manufactured as a feasibility test by PIM process. Average grain size and porosity of Li2TiO3 pebbles were less than 1.00 µm and about 39.7 % with almost open pore type, respectively. The high-intensity D-T fusion neutron source (HINEG-CAS) was employed at the Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences (INEST, CAS) to conduct a neutron irradiation experiments. The total number of fusion neutrons was about 1.104 x 1015 n. Tritium release experiment from the irradiated pebbles was also conducted using the tritium release system from room temperature up to 800 oC. The total radioactivity of release tritium from the irradiated Li2TiO3 pebbles with 274.1 g in weight was about 1866.4 Bq. Tritium release behavior including the ratio between HTO and HT at elevated temperature was preliminarily investigated in this study.

Speaker: Yi-Hyun Park

7 A new perspective on hydrogen isotope permeation behavior of LiAlO2 layer

Long Wang, Fantao Meng, Zhihao Hong, Baoping Gong, Fengchao Zhao, Qixiang Cao
Southwestern Institute of Physics, Chengdu 610225, P.R. China
Formation of a LiAlO2 layer is inevitable once solid-state reaction occurred between Al2O3 tritium permeation barrier and lithium ceramics. In this study, effect of LiAlO2 layer on the hydrogen isotope permeation behavior was first investigated. LiAlO2 coating were prepared on CLF-1 steel substrates by high-temperature lithium infiltration. The formation process and hydrogen isotope permeation behavior of LiAlO2 coatings were systematically analyzed through experimental and simulation approaches. The results demonstrate that LiAlO2 exhibits inward growth over time, forming a stable LiAlO2 -Al interface after 7 days. Deuterium permeation tests revealed that the LiAlO2 coating effectively resists deuterium permeation and diffusion. Notably, as the LiAlO2 coatings thickens, both the permeability and diffusion rates decrease progressively. Computational simulations further corroborated these findings, attributing the resistance to the low H₂ adsorption energy and high dissociation barrier of the LiAlO2 surface.

Speaker: Long Wang

8 Features of experiments to study the generation and release of tritium from lithium-containing ceramics in conditions of vacuum pumping.

I.Ye. Kenzhina1,2, T.V. Kulsartov1, A.A. Shaimerdenov1,2, Ye.V. Chikhray1, S.K. Askerbekov1,2, M. Aitkulov2, Zh.A. Zaurbekova1
1Satbayev University, Almaty, Kazakhstan
2Institute of Nuclear Physics, Almaty, Kazakhstan
e-mail: kenzhina@physics.kz
The design of future fusion reactors involves the generation of tritium inside a breeder blanket. One of the most promising materials for the breeder blanket today is lithium ceramics. Tritium is formed in lithium under neutron irradiation by the reaction 6Li(n,α)T. This tritium is then extracted from the blanket by purge gas and returned to the fusion area, realizing the concept of a closed fusion cycle. Irradiation under fission reactor conditions is still one of the few available methods for in-situ estimation of tritium generation and release parameters from lithium-containing materials. In this work, based on a series of reactor experiments performed with two-phase lithium ceramics and LMT ceramics, a number of advantages of using the technique to investigate the generation and release of tritium when conducting experiments under vacuum pumping conditions of the samples under study have been identified. The main advantages of this method include:
• It is possible to fully control the composition of the gas phase in the chamber with samples, which greatly facilitates the analysis of possible reactions on the ceramic surface associated with the release of both tritium-containing molecules and helium;
• possibility of conducting experiments to study the release of tritium from lithium ceramics at a given change in the composition of gases in the chamber with samples;
• high sensitivity and practically no delay in recording changes in the gas phase composition (including tritium-containing molecules and helium) when changing various experimental parameters (such as sample temperature; reactor power, etc.)
A number of methodological complexities of reactor experiments with lithium ceramics under conditions of vacuum pumping of the studied samples are also analyzed:
• in particular, it was found that there are some limitations on the lower limit of the investigated temperatures of the ceramics, which is caused by the low thermal conductivity of pebble bed;
• also, when the samples are arranged in several layers in the pebble bed, there is a noticeable temperature gradient across the filling, which is difficult to measure and can often be estimated only by calculation;
• it is better to use high-resolution mass spectrometers in the 4-mass region to separate the signal for HT and helium molecules.

Speaker: I Ye Kenzhina

9 Tritium Transfer Behavior from Neutron-Irradiated LiAlO 2 to Zirconium by Heating in a Sealed Quartz Tube

Hiroki Isogawaa, Kazunari Katayamaa, Rin Ganahab, Hideaki Matsuurac
aDepartment of Advanced Energy Engineering Science, Kyushu University,
6-1 Kasugakouen Kasuga-shi Fukuoka, 816-0811, Japan
bSchool of Engineering, Kyushu University,
744 Motooka Nishi-ku Fukuoka-shi Fukuoka, 319-0395, Japan
bcDepartment of Applied Quantum Physics and Nuclear Engineering, Kyushu University,
744 Motooka Nishi-ku Fukuoka-shi Fukuoka, 319-0395, Japan

To ensure sufficient tritium inventory for the start-up phase of D-T fusion reactors, we have proposed a method of tritium production by the ⁶Li(n, α)T reaction in high-temperature gas-cooled reactors (HTGR). A key challenge is minimizing tritium loss under high-temperature conditions. While Li₂TiO₃ and Li₄SiO₄ are promising materials for blankets in fusion reactors, LiAlO₂—known for its excellent chemical stability at high temperatures—has been selected as a primary candidate for tritium production in HTGRs. A potential solution involves encapsulating LiAlO₂ with zirconium (Zr) in an Al₂O₃ container, but the tritium behavior in such composite systems is not yet fully understood.
In this study, we investigated tritium transfer behavior from LiAlO2 to Zr under heating conditions. LiAlO2 and Ni coated Zr were sealed in a quartz tube and irradiated by neutrons in the JRR-3 reactor. After irradiation, the following 3 experiments were conducted to evaluate the extent of tritium transfer from LiAlO2 to Zr.
・Run 1: LiAlO₂ powder + Ni-coated Zr spheres, heated to effectively 700 °C
・Run 2: LiAlO₂ pebble + Ni-coated Zr spheres, heated to 900 °C
・Run 3: LiAlO₂ powder + Ni-coated Zr spheres, heated to 1000 °C
The sealed samples were pre-heated to the target temperature at 30 °C/min and held for 60 minutes. In an Ar-filled glovebox, Zr spheres and LiAlO₂ were separated, and then individually reheated to either 900 °C or 1000 °C at 5 °C/min. Tritium release rates over time were quantified using sequential water bubblers: tritiated water vapor (HTO) was collected in the first bubbler, and gaseous tritium (HT), after oxidation, in the second. The tritium retention ratios in LiAlO₂, Zr, and other components were as follows:
・Run 1: LiAlO₂ powder: Ni-coated Zr : others = 94.0 : 4.3 : 1.7
・Run 2: LiAlO₂ pebble : Ni-coated Zr : others = 89.0 : 9.9 : 1.1
・Run 3: LiAlO₂ powder : Ni-coated Zr : others = 95.0 : 4.7 : 0.3
These results indicate that more than 90% of the tritium was retained within the LiAlO₂-Zr system after heating above 700 °C. The slightly lower retention in Run 2 may be attributed to the sintering of LiAlO₂ pebbles, which likely led to crystal grain growth and reduced tritium diffusivity. The "others" category represents tritium lost via permeation through the quartz tube during pre-heating and tritium released into the Ar atmosphere during the post-heating quartz breakage process.

Speaker: Hiroki Isogawa

10 Investigation of hydrogen isotope release behavior in long-term annealed tritium breeding pebbles

Qiang Qi1,2*, Shouxi Gu1, Yingchun Zhang3, Hai-Shan Zhou1,2, Guang-Nan Luo1,2

1Institute of Plasma Physics, Hefei Institutes of Physical Sciences, Chinese Academy of Sciences, Hefei, 230031, China
2Science Island Branch, Graduate School of USTC, Hefei, 230036, China
3School of Materials Science and Engineering, University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083, PR China

Lithium-based ceramic tritium breeders are key materials for fusion blankets, producing tritium by neutron-induced lithium transmutation. These pebbles must endure long-term high temperature service in fusion conditions, and their microstructural and property changes impact safety and tritium self-sufficiency.
This work examines the microstructural, phase, and compositional changes in Li4SiO4 pebbles after annealing at 900°C in a 0.1 H2/He atmosphere for various durations, and their effects on deuterium release behavior. Initially, Li4SiO4 pebbles contain Li4SiO4, Li2SiO3, and Li2CO3 phases. Annealing eliminates the Li2CO3 phase, but minor reformation occurs due to CO2 absorption during sampling and storage. Significant oxygen and lithium loss during annealing is undesirable, as it affects tritium breeding and may cause safety issues by blocking gas flow and corroding structural materials. Grain growth and pore coalescence also occur, degrading mechanical performance.
Surprisingly, despite expectations, deuterium release behavior does not deteriorate after 1000 h of annealing. The transformation of closed pores to open pores likely promotes the shift of deuterium release peaks to lower temperatures. The distinct deuterium release behavior between virgin and annealed pebbles is mainly due to carbonate impurities. The high-temperature HDO release peak at 665°C in virgin pebbles is attributed to Li2CO3 decomposition. Drying Li4SiO4 pebbles at 700°C for several hours to remove carbonate impurities is recommended before application. Further neutron irradiation experiments are needed to evaluate tritium release behavior.

Speaker: Qiang Qi

3:10 PM Coffee break

Irradiations

Convener: Julia Leys

11 Investigating the Effect of High Energy Ion Irradiation and High Temperatures on the Mechanical Properties and Phase Stability of the Highly Lithium Dense ‘Octalithium Ceramics’ (Li8MO6, M = Zr, Sn, Pb and Ce) for Tritium Breeding

Pedr Charlesworth1*, Ben Phoenix2, Samuel Murphy3, Mohamad Abdallah4, David Kingham4,
David Armstrong1, Chris Grovenor1
1Department of Materials, University of Oxford, U.K.
2Department of Physics and Astronomy, University of Birmingham, U.K.
3 Engineering Department, Lancaster University, U.K
4Tokamak Energy, Abingdon, U.K
In order to most efficiently produce tritium from a high energy neutronic reaction,
lithium dense tritium breeding materials (TBMs) are required. TBMs must operate under
high temperatures and neutron radiation, whilst producing extractable tritium and being
compatible with the surrounding materials. Ceramic TBMs offer material compatibility and
do not suffer from magnetohydrodynamic (MHD) effects, however, traditionally they have
lower tritium breeding ratios (TBRs) in addition to concerns over radiation damage.
With the current industrial interest in spherical tokamak arrangements with less
space for TBMs, materials with higher TBRs are required. Neutronics simulations suggest
that the octalithium compounds, with their high lithium densities, offer significantly higher
TBRs than Li4SiO4 and Li2TiO3 which are designated for use in ITER – however most of these
compounds lack basic physical data (melting points, phase stability, mechanical properties)
and none have been subject to micro mechanical and ion irradiation testing.
This work presents the mechanical properties (Youngs modulus, hardness and
fracture toughness) of dense octalithium ceramics (Li8MO6, M = Zr, Pb, Sn and Ce) from
nanoindentation, how these experimental values correspond with those predicted using
density functional theory modelling (DFT), and the impact of high energy (12 MeV, 1e17cm-2)
He ion irradiation on these properties. Further we examine how the octalithiums will
perform in the hostile environment of a future reactor, by exploring the phase stability at
high temperatures (500°C, 700°C and 900°C) using X-ray diffraction and mass loss.

Speaker: Pedr Charlesworth

12 Multiple irradiation campaigns for the testing of solid breeder specimens

Chase Taylor, Pattrick Calderoni, Xiao-Ying Yu

Speaker: Chase Taylor

13 Neutronics analysis of the ceramic breeder blanket materials irradiation in the IFMIF-DONES test modules

Ali Abou-Sena, Frederik Arbeiter, Santiago Becerril, Achim Kupferschmitt, Dieter Leichtle, Jin Hun Park, Yuefeng Qiu, Arkady Serikov, Guangming Zhou

Speaker: Dr Arkady Serikov (Karlsruhe Institute of Technology (KIT))

14 Development and Application of Machine Learning Potential for Lithium Titanate for Radiation Damage Simulations

Donggyu Lee, Takuji Oda
Seoul National University

Lithium titanate (Li2TiO3) is a promising ceramic-type tritium breeder material for nuclear fusion reactors. While extensive experimental studies have been conducted to investigate its properties, the understanding of its behavior under radiation environments remains limited. Molecular dynamics (MD) simulations provide crucial insights into defect formation and evolution and their impact on material properties. Previous MD studies on Li2TiO3 have primarily relied on a Buckingham-type empirical two-body potential combined with the universal ZBL potential.
In recent years, machine learning (ML) potentials have emerged as a game changer in atomistic simulations, achieving first-principles accuracy while significantly reducing computational cost. We have recently developed an ML potential for Li2TiO3 based on the moment tensor potential framework. Our evaluation shows that the ML potential outperforms the conventional Buckingham-type potential in reproducing bulk material properties, such as heat capacity and thermal conductivity, and is in better agreement with DFT calculations and experimental data.
In this presentation, we will demonstrate the applicability of our ML potential in analyzing defect systems. Our ML potential can assist in geometry optimization tasks by providing approximate defect structures and energies at a lower computational cost than first-principles calculations, and allows us to thoroughly perform geometry optimization, which is a formidable task for first-principles calculations because the flat potential landscape of Li2TiO3 causes slow convergence. Furthermore, the accuracy and efficiency of ML potentials allow the exploration of various defect compositions and configurations. These results illustrate how ML potentials facilitate a comprehensive analysis of defect structure and stability.

Speaker: Donggyu Lee

15 Annealing self-healing behavior of ion-irradiated tritium breeders and the compatibility with RAFM steel

Shouxi Gu1, Qiang Qi1,2, Hai-Shan Zhou1,2
1Institute of Plasma Physics, Hefei Institutes of Physical Sciences, Chinese Academy of Sciences, Hefei, 230031, China
2Science Island Branch, Graduate School of USTC, Hefei, 230036, China

The neutron energy spectrum in solid breeder blankets spans from 10−8 to 14 MeV, generating primary knock-on atoms (PKAs) with energies ranging from 10−4 to several MeV within the breeder material. These PKAs induced by neutron irradiation trigger cascade collision processes, producing substantial displacement damage that threatens irradiation stability. 5 MeV Si ion irradiation was adopted to bombard the mixed-phase Li2TiO3-Li4SiO4 pebbles. Electron spin resonance (ESR) analysis revealed irradiation-induced vacancy-type defects, including oxygen vacancy defects (E’-centers) and non-bridging oxygen hole centers (NBOHCs). Raman spectroscopy demonstrated severe damage to the fundamental structural units of the material, resulting in broadened spectral features with no distinct peaks. Annealing experiments at 300°C and 650°C revealed self-healing behavior as the annealing temperature increased. The damaged structural units gradually recovered, with a significant recovery phenomenon observed in Raman spectra after high-temperature annealing. This suggests that thermal treatment effectively mitigates irradiation-induced lattice disorder, highlighting the material’s potential for self-repair under operational conditions in fusion environments.
Neutron irradiation will generate gaseous transmutation products, H and He, through (n, xH) and (n, xHe) reactions in RAFM steel. These energetic gaseous particles will induce radiation defects in steels. Helium ions as insoluble gases will form voids and bubbles in irradiated steels, which will alter the corrosion process of steels. To study the influence of transmutation-produced helium on the corrosion behavior of CLF-1 steel by lithium ceramic tritium breeders, 100 keV helium ion irradiation of CLF-1 steel was conducted. Corrosion experiments in Li4SiO4 powder at 550 ℃ were conducted for helium ion irradiated CLF-1 steel. The GIXRD of the non-irradiated and helium ion irradiated CLF-1 steel samples after 10 h of corrosion in Li4SiO4 powders, reveals that the corrosion product is the Fe3O4 phase. After irradiation, the diffraction peak intensity of the Fe3O4 phase decreased significantly. The post-irradiation TEM image discovers a separation between the corrosion layer interface and the substrate, with abundant helium bubbles observed near the interface on the substrate side, indicating helium aggregation and growth during corrosion. The sizes of helium bubbles in post-irradiated CLF-1 steel before and after corrosion are measured, respectively. After corrosion, the helium bubbles grew to approximately 2–3 nm, significantly larger than those in the uncorroded CLF-1 steel. Based on these findings, the corrosion layer spallation mechanism of helium-irradiated CLF-1 steel has been proposed.

Speaker: Shouxi Gu

5:50 PM [OPTIONAL] Meet at bus stop for BeWS dinner

Those who registered to attend the BeWS dinner, please come to a meeting point for bus. (The fee for the workshop conference dinner is included in the registration fee of BeWS). The meeting point will be within Uji campus Kyoto University, which will be announced soon.

Tue, October 7

CBBI/BeWS joint session 3

Convener: Mu-Young Ahn

16 [BeWS] Rockland Resources and Beryllium as a Critical Mineral in Clean Energy Research

Speaker: Chris DORN

17 Impact of various HCPB blanket modifications for the new EU-DEMO baseline on tritium breeding ratio and neutron shielding performance

Roman Afanasenko1, D. Leichtle1, F. Hernández1, J.H. Park1, P. Pereslavtsev2, G. Zhou1
1Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen, Karlsruhe, Germany
2EUROFUSION Programme Management Unit, Fusion Technology Department - DEMO Central Team, Boltzmannstrasse 2, 85748 Garching, Germany

The Helium-Cooled Pebble Bed (HCPB) breeding blanket is one of the leading candidates for the new DEMO baseline, requiring efficient tritium breeding and strong neutron shielding to ensure reactor sustainability and structural integrity. This study explores innovative modifications to the configuration of the HCPB breeder blanket. The effects of changes in the breeder spherical layer geometry, neutron multiplier distribution, and cooling channel location on the tritium breeding ratio (TBR) and neutron flux decay are investigated. This work provides a first preliminary assessment of HCPB availability for the new DEMO baseline, balancing gains in tritium breeding against shielding and material challenges.
Using Monte Carlo neutron transport simulations, the neutronic performance of each variant was evaluated under DEMO-relevant conditions. These analyses demonstrate that the HCPB concept of the breeding blanket represents a highly effective and robust tritium breeding system for the DEMO tokamak. The design achieves a high tritium breeding ratio (TBR) while maintaining adequate shielding to protect critical components, even with localized adjustments to compensate for changes in neutron flux distribution. The results confirm that the HCPB blanket ensures sufficient tritium self-sufficiency and structural integrity under operational loads.

Speaker: Roman Fanasenko

10:10 AM Coffee break

CBBI/BeWS joint session 4

Chair: Regina Knitter (KIT)

Convener: Regina Knitter

18 [CBBI] Neutronics experiments on the mock-ups of WCCB and COOL blanket under D-T neutron irradiation conditions

Qingjun Zhu a, Qiankun Shao a, Wuhui Chen a, Jie Bao b, Hua Dua a, Songlin Liu a
a. Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031, China
b. Department of Nuclear Physics, China Institute of Atomic Energy, Beijing 102413, China

Abstract:The tritium production performance of the fusion blanket reaches the expected index, which is the primary prerequisite for the fusion reactor to realize tritium self-sustaining operation. The demonstration of this premise has a decisive impact on the design of fusion blanket. In order to verify the reliability of the neutronics design tool, i.e. MCNP and FNEDL databases, the mock-up of water-cooled ceramic breeder (WCCB) blanket and supercritical carbon dioxide cooled Lithium-Lead (COOL) blanket were developed, and the neutronics experiments were carried out using DT neutron generator, the simulated and experimental values (C/E) of neutronics parameters represented by tritium production rate (TPR) were analyzed. In order to realize the consistency of TPR simulation and measurement in neutron source terms, the real-time source item analysis program based on the associated particle spectra for DT neutron generator source item was developed. For TPR validation, two independent technologies were used, including the Li2TiO3/Li2CO3 pellets as off-line TPR detectors, and the developed lithium glass detectors as on-line TPR detectors. The TPR validation was complemented by evaluating the neutron-induced reaction rates of activation foils. The experiments are in good agreement with the simulations in the central axis of mock-up, while it was indicated that there are unexpected scattered neutrons in the edge region of the mock-up, resulting in the higher TPR than expected.

Speaker: Qingjun Zhu

19 [BeWS] Permeation-against-vacuum extraction of hydrogen isotopes in a forced convection FLiBe loop

Speaker: Alexander HINZ

20 [CBBI] Overview of activities at INP to irradiation testing of ceramic breeder

A.Shaimerdenov, T.Kulsartov, I.Kenzhina, Zh.Zaurbekova, Sh.Gizatulin, S.Askerbekov

The Institute of Nuclear Physics, 1 Ibragimov st., Almaty, Kazakhstan

This article provides an overview of irradiation activities on the WWR-K reactor to test ceramic blanket materials for a fusion reactor. It describes the experimental capabilities of the WWR-K reactor, the tested materials, the conditions of the irradiation test, and the results of an in-reactor study of ceramic materials.
The irradiation tests demonstrated the processes and features of tritium generation and release from ceramic materials under different conditions (temperature, neutron fluence, environment). The experimental data obtained have a practical contribution to ceramic breeder material development for ITER and DEMO.

Speaker: A. Shaimerdenov

12:10 PM Lunch break

Characterizations

Convener: Yihyun Park

21 Influence of lithium precursor deficiency and excess during solid-state synthesis on radiation-induced effects in biphasic Li4SiO4-Li2TiO3 ceramics

Arturs Zarins1,2, Laura Dace Pakalniete1, Annija Liepkalne1, Liga Avotina1, Artis Kons3, Guna Krieke4, Jekabs Cirulis4, Andris Antuzevics4
1 – Institute of Chemical Physics, Faculty of Science and Technology, University of Latvia, Riga, Latvia; 2 – Department of Environment and Technologies, Faculty of Natural Sciences and Healthcare, Daugavpils University, Daugavpils, Latvia; 3 – Department of Chemistry, Faculty of Medicine and Life Sciences, University of Latvia, Riga, Latvia; 4 – Institute of Solid State Physics, University of Latvia, Riga, Latvia

Biphasic lithium orthosilicate (Li4SiO4) – lithium metatitanate (Li2TiO3) ceramics are being developed and tested worldwide as a potential material for tritium breeding in thermonuclear fusion reactors. In the present work, the influence of lithium precursor deficiency and excess during solid-state synthesis on radiation-induced effects in biphasic Li4SiO4-Li2TiO3 ceramics is investigated. The two-phase Li4SiO4-Li2TiO3 powder samples with various contents of lithium precursor were prepared and characterised using different physico-chemical methods, e.g., X-ray diffraction, attenuated total reflectance – Fourier transform infrared spectroscopy, etc. For comparison purposes, the single-phase Li4SiO4 and Li2TiO3 powder samples were prepared using similar approach. Rietveld analysis of high-resolution X-ray diffraction patterns was employed to determine the potential changes of crystal lattice parameters and atomic positions. The formed and accumulated paramagnetic radiation-induced defect centres in the prepared samples after irradiation with X-rays were studied using electron paramagnetic resonance spectroscopy. The obtained results show that the deficiency and excess of the lithium precursor considerably influence the phase composition of the biphasic Li4SiO4-Li2TiO3 samples. Although the types of paramagnetic centres generated in the prepared samples are similar, their relative intensities vary depending on the content of lithium precursor.
This work has been carried out within the framework of the Latvian Council of Science project No. LZP-2024/1-0162 “Influence of stoichiometry on radiation-induced effects in advanced two-phase functional materials for future thermonuclear fusion reactors”.

Speaker: Arturs Zarins

22 Synthesis, Characterization, and Testing of High-Lithium-Density Composite Breeders

Michael Moorehead, Priyanshi Agrawal, Malachi Nelson, Chase Taylor, Wei Tang, Timothy Yoder, Mario Daniel Matos II, Jorgen Rufner, Pierre-Clement Simon, Lin Yang
Idaho National Laboratory

Solid tritium breeder materials must first and foremost have sufficiently high concentrations of lithium to enable a plant-scale tritium breeding ratio greater than 1:1. However, in addition to lithium content, such breeder materials must also meet other performance metrics including high tritium release rates, thermal conductivities, and irradiation damage tolerance. Perhaps most importantly, tritium breeders must maintain their mechanical integrity during reactor operation so as to avoid degradation which can jeopardize the functionality of the tritium breeder blanket module, which in most designs takes the form of a pebble bed geometry. Unfortunately, the mechanical robustness of most lithium-bearing ceramics under investigation for fusion applications is often inversely related to the lithium atom density. For example, a material such as lithium oxide (Li2O), which has one of the highest lithium atom densities, has a much lower mechanical splitting strength than lithium metatitanate (Li2TiO3), though Li2TiO3 has less than half the lithium atom density of Li2O.
This work seeks to provide an alternative to monolithic ceramic tritium breeders, in the form of metal-reinforced composite tritium breeders. Specifically, composite tritium breeders have been synthesized combining Li2O with various ferrous metal reinforcements via electric field assisted sintering (EFAS), also known as spark plasma sintering (SPS). As the metal reinforcement content is increased, metallic networks are observed, via electron microscopy and X-ray computed tomography, to form throughout the composite material. Through destructive mechanical testing, even dilute metal reinforcement loading enables drastic mechanical strength improvements over pure Li2O while higher loadings give rise to quasi-ductile behavior and higher ultimate strengths than Li2TiO3 – while still maintaining a higher density of lithium atoms than Li2TiO3 and many other breeder candidates. In addition to microstructural characterization and mechanical testing, thermal property measurements and hydrogen permeability testing are underway to further assess the suitability of such composites for fusion reactor applications.

Speaker: Michael Moorehead

23 Corrosion behavior of F82H by LTZO ceramic breeder pebbles

Keisuke Mukai,1,2*, Kosuke Kataoka3, Juro Yagi3, Motoki Nakajima4, Jae-Hwan Kim4, Takashi Nozawa4
1　National Institute for Fusion Science, National Institutes of Natural Science, Toki, Gufu 509-5292, Japan
2　The Graduate University for Advanced Studies, SOKENDAI, Toki, Gifu 509-5292, Japan
3　Graduate School of Energy Science, Kyoto University, Uji, Kyoto 611-0011, Japan
4　National Institutes for Quantum Science and Technology, Rokkasho, Aomori, Japan

In ceramic breeding blanket, reduced activation ferritic/martensitic steel (RAFM) structural steel is corroded by vapor gas released from tritium breeder (Li-containing ceramic pebbles), resulting in a significant reduction of the fatigue lifetime. To understand the corrosion kinetics and the effect of moisture absorbed in the breeder pebbles, the corrosion test of F82H by LTZO (Li2+xTiO3+y solid solution with 20 wt% Li2ZrO3) pebbles were conducted at 773–998 K. Microstructural analysis was conducted using scanning electron microscope (SEM) and energy dispersive spectroscopy (EDS). Element depth profile and structural analysis using glow discharge optical emission spectroscopy (GD-OES) and X-ray diffraction (XRD) identified the corrosion products has cubic, spinel, and rhombohedral structures with the compositions of Li–TM–O (TM: transition element in F82H). In the long-term compatibility test for up to 672 h, the growth of the corrosion layer thickness followed a parabolic law at 833 K, yielding apparent diffusion coefficient of 6.95 × 10–13 cm2/s. A rapid growth of the corrosion layers was observed at 993 K after 380 h which could be triggered by failure of the protective layer. A comparison of diffusion coefficients with the reported data indicated predominant effects of temperature and impurity concentrations in the sweep gas on the corrosion, while the composition and shape of ceramic breeder had minor influences. The apparent diffusion coefficients obtained in this work were similar　to those of gas-oxidation of Fe-Cr alloy. The estimated thickness of the corrosion layer was only 5.8 μm at 623 K after 2 years. The effect of corrosion on fatigue behavior of F82H steel will be assessed in our future study.

Speaker: Keisuke Mukai

24 Effect of Helium Behavior on Hydrogen Isotope Release in Li2TiO3

Shurui Shang1,2, Qiang Qi1,2*, Shouxi Gu1, Hai-Shan Zhou1,2, Guang-Nan Luo1,2

1Institute of Plasma Physics, Hefei Institutes of Physical Sciences, Chinese Academy of Sciences, Hefei, 230031, China
2Science Island Branch, Graduate School of USTC, Hefei, 230036, China

Lithium-based ceramic tritium breeders serve as critical components in fusion blanket systems. Under neutron irradiation conditions, neutron-induced transmutation reactions that produce tritium (³H) simultaneously generate helium-4 (⁴He) atoms in a 1:1 stoichiometric ratio. Owing to its characteristically low solubility in ceramic matrices, helium readily interacts with irradiation-induced vacancies. This interaction may drive void nucleation, followed by helium accumulation within these voids to form helium bubbles. Critically, such defects induced by helium behavior may act as trapping sites for tritium, thereby altering its release dynamics.The changes of phase, microstructure, chemical states and its effects on release behavior of hydrogen isotopes in helium and deuterium irradiated Li2TiO3 have been investigated. SRIM-2013 simulations revealed that 100 keV helium (He) ion irradiation induced lattice defects distributed up to 500 nm beneath the surface, with a peak damage level of 2.7 displacements per atom (DPA). Following helium ion irradiation, the characteristic Ti2p peaks at 458.2 eV and 464.0 eV were quenched. Following 50 keV deuterium (D) ion irradiation, the helium-containing microstructure of Li₂TiO₃ was characterized by transmission electron microscopy (TEM), while the deuterium release kinetics were analyzed via thermal desorption spectroscopy (TDS).

Speaker: Shurui Shang

25 The enhanced sintering and mechanical properties of Li4SiO4 by the introduction of a secondary phase

Qilai Zhou a, Yifu Xu a, Kaixuan Zhu a
a School of Materials Science and Engineering, Wuhan University of Technology, Wuhan 430070, China

Abstract:
Lithium orthosilicate (Li4SiO4) is considered as a promising solid tritium breeding material for fusion reactors due to its high lithium density and subsequent high tritium generation rate. However, its limited sinterability and inadequate mechanical properties present significant challenges. This study investigates the addition of neutron multiplier elements, Pb and Zr, to improve the sintering process and microstructure of Li4SiO4 ceramics, aiming at enhanced densification and improved mechanical performance. Pb was introduced into the precursor solution in the form of lead nitrate, Pb(NO3)2, and Li4SiO4-Pb ceramic powders were synthesized via a microwave-assisted solution combustion method. It was found that Pb existed as PbO in the ceramic. During sintering, PbO melted and accumulated at the Li4SiO4 grain boundaries, enhancing densification through liquid-phase sintering and improving mechanical properties. The Li4SiO4-10 wt.% Pb ceramic pebbles sintered at 1173 K achieved a relative density of 96.4% and a crush load of 106.3 N. However, PbO also promoted grain growth. In contrast, the addition of 10 wt.% ZrO2 led to the formation of Li2ZrO3, which induced a pinning effect, effectively refining the grain size. The ceramic pebbles exhibited a relative density of 86.2% and a crush load of 40.8 N. Furthermore, bending strength tests on ceramic bodies demonstrated a positive correlation between bending strength and densification. These findings suggest that Pb and Zr additions can effectively modify the microstructure and mechanical properties of Li4SiO4 ceramics, providing valuable insights for their application in fusion reactors.

Speaker: Qilai Zhou (Wuhan University of Technology)

3:10 PM Coffee break

Synthesis & Enrichment

Convener: Pedr Charlesworth

26 Results from the Continuous Operation of the KALOS Process

Oliver Leysa, Julia Leysa, Regina Knittera
a. Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM), Karlsruhe, Germany

Abstract:　 The KALOS process is used at the Karlsruhe Institute of Technology for the production of Advanced Ceramic Breeder (ACB) pebbles. These pebbles, composed of lithium orthosilicate with a strengthening phase of lithium metatitanate, are regarded as the EU reference solid breeding material and will be featured in the ITER HCCP, as well as DEMO blankets. The melt-based process involves heating synthesis powders in a platinum alloy crucible inside a furnace to form a melt. Pressure is applied to the crucible to force the melt through a small nozzle, thereby forming a laminar jet, which breaks up into droplets that are then solidified in a cooling tower to form pebbles. After production, a comprehensive standardised characterisation is performed on each batch.
In order to provide the roughly 90 kg of ACB pebbles that will be required for an ITER test blanket module, the process recently underwent a significant upgrade, where the main goal was to increase the capacity of the process, while maintaining the high quality of the pebbles. In the past, the maximum batch size was limited by the volume of the melting crucible, which in turn was limited by the useable space inside the process furnace. To address this problem, the process has been converted from a batch operation to a process where the synthesis powders can be continuously fed into the melting crucible. The changes have resulted in a theoretical maximum production capacity of 10 kg in a working day.
This work will look at the first production results for the first batches produced using the new process set-up. This includes an examination of various aspects of the process stability and reliability, including critical variables such as the process pressure or temperatures at the nozzle and in the cooling tower. The characterisation of the pebbles focuses on the main pebble properties, including the pebble sizes and distribution, the mechanical strength, the porosity, as well as an analysis of the phases and chemistry.
The findings confirm the reliability and consistency of the upgraded process, as well as the high quality of the produced pebbles. This shows that not only will the KALOS process be able to produce the pebbles required for testing in ITER, but that it also demonstrates that it is possible to upscale the technology to an industrial scale.

Speaker: Oliver Leys (Karlsruhe Institute of Technology)

27 Powder Synthesis with Core-shell Structure for Tritium Breeder

Young Ah Park, Yi-Hyun Park
Korea Institute of Fusion Energy (KFE), Korea

Lithium orthosilicate (Li4SiO4) and Lithium metatitanate (Li2TiO3) have been studied as promising tritium breeder for fusion energy. Li4SiO4 has a lithium density of 0.51g/cm3, which is higher than the lithium density of Li2TiO3 (0.43g/cm3), resulting in higher tritium production efficiency. However, Li4SiO4 is prone to decomposition reaction at high temperature, resulting in low mechanical-chemical stability at the operation temperature. Recently, core-shell type Li4SiO4-Li2TiO3 pebble, coated with Li2TiO3 on the surface of Li4SiO4 pebble, has been developed to enhance the low mechanical-chemical stability. (Core: Li4SiO4, Shell: Li2TiO3) It is usually prepared in two steps: fabrication of Li4SiO4 pebbles and coating with Li2TiO3. This Li4SiO4-Li2TiO3 pebble fabrication method is not suitable from the point of view of mass production for breeding blanket. In addition, the difference in shrinkage rates between Li4SiO4 and Li2TiO3 provides a potential for separation of the interface between core and coating during fabrication and operation.
In this study, we aim to synthesize Li4SiO4@Li2TiO3 powders with a core-shell structure to improve the low thermal stability of Li4SiO4 and simultaneously prepare pebbles that can be mass-produced. (Core: Li4SiO4, Shell: Li2TiO3) Li4SiO4@Li2TiO3 powders were synthesized at nanoscale using hydrolysis and syn-lithiation methods. The detailed preparation process and the properties of synthesized Li4SiO4@Li2TiO3 powder will be described in this presentation.

Speaker: Young Ah Park

28 Enrichment of Lithium-6 using Electrodialysis with Ionic Conductor Membrane

Kenji Morita, Kodai Ishii, Makiko Sekiya, Yumiko Tanaka, Tsuyoshi Hoshino
Rokkasho Institute for Fusion Energy, National Institutes for Quantum Science and Technology (QST), Rokkasho, Aomori 039-0212, Japan.
Keywords: Lithium-6 Enrichment, Ionic Conductor, Electrodialysis, Breeding blanket
In fusion power generation, achieving tritium self-sufficiency necessitates a sufficient ratio of lithium-6 (6Li) in the breeding blanket materials. While Lithium is abundant in nature, natural abundance of 6Li in lithium is only 7.6%. Therefore, isotope separation technology is crucial to meet the required 6Li ratio for fusion reactors. The practically applied lithium isotope separation method is the amalgam method using mercury, which raises environmental concerns, making the development of alternative technologies desirable.
At QST, research and development are underway for the extraction of high-purity lithium and the enrichment of 6Li using electrodialysis with a lithium-ion conducting solid electrolyte membrane (Lithium Separation Method by Ionic Conductor: LiSMIC). For the enrichment of 6Li, this technology exploits the mass-dependent difference in the diffusion rates of lithium ions within the solid electrolyte. According to the transition state theory, the ratio of the diffusion coefficients for 6Li and 7Li is given by the square root of the inverse mass ratio, approximately 1.08, which characterizes the separation and enrichment performance. By applying a voltage across the membrane, lithium ions migrate from the anode to the cathode under the influence of the electric field. Owing to its slightly higher mobility, 6Li is expected to migrate faster, leading to its enrichment on the cathode side. Consequently, the characteristics of the applied voltage can influence the separation performance of 6Li.
This presentation will discuss the impact of applied voltage characteristics on the separation factor and the migration rate of lithium, highlighting the trade-off relationship between these two parameters. Given that a 6Li enrichment level of 90% is required for the DEMO reactor, a multi-stage LiSMIC system will be essential for achieving this target. Therefore, cascade theory will be applied to the obtained experimental data to estimate the scale of a potential production facility.

Speaker: Kenji Morita

29 Improvement of lithium isotope enrichment ratio and enrichment rate by electrochemical active control of adsorbed ions in intermittent voltage electrodialysis

Kiyoto Shin-mura1, Shumpei Iwasaki1, Kazuya Sasaki1
1Hirosaki University

Current major plans for thermonuclear power generation use the nuclear fusion reaction of deuterium and tritium. Tritium, which is almost absent in nature, must be generated in a fusion reactor by the fission reaction of lithium 6 isotope (6Li) with neutrons. For this purpose, technology for enriching the 6Li isotope is required. In electrodialysis using a solid electrolyte membrane, lithium isotopes diffuse at a rate proportional to the inverse of the square root of their mass number. As a result, theoretically, an isotope enrichment ratio of approximately 1.08 occurs inside the solid electrolyte. Kunugi et al. experimentally demonstrated a 6Li isotope enrichment ratio of 1.08 (Solid State Ionics, 122 (1999), pp. 35-39). However, this mass effect is limited to a short voltage application time. On the other hand, we have clarified that intermittent voltage application sustains a high 6Li isotope enrichment ratio (Journal of the Ceramic Society of Japan, 126 (2018), pp. 331-335). Furthermore, by lowering the temperature of the electrolyte membrane, a large isotope enrichment ratio exceeding 1.08 was achieved through quantum effects (Fusion Engineering and Design, 171 (2021), 112577). However, these methods require long periods of interruption in voltage application, resulting in low lithium transfer rates and a need for improvement. In this study, we investigate the effect of active electrochemical control of the isotope ratio and amount of lithium adsorbed on the electrolyte membrane surface, and clarify that a larger isotope enrichment ratio and 6Li transfer rate can be achieved.

Speaker: Shumpei Iwasaki

Wed, October 8

Modeling & Analysis

Chair: Kenji Morita (QST)

Convener: Kenji Morita

30 Modelling transport of dust particles, and fragmentation estimates in the HCPB-BB of DEMO

Dario Passafiume1, Marc Kamlah1
1Karlsruhe Institute of Technology, Institute for Applied Materials, Karlsruhe, Germany

The European solid breeder blanket concept Helium Cooled Pebble Bed (HCPB) uses advanced lithium ceramic breeder (ACB) material in the form of pebbles. During the HCPB breeder blanket operation, a fragmentation of the pebbles due to thermomechanical loads can occur, as well as the formation of dust. This dust represents a safety issue, as it can block purge gas paths inside the HCPB or it can be transported with the purge gas from the pebble bed into the tritium extraction system (TES) and accumulate there, especially in filters. Therefore, it is crucial to understand the ACB pebbles’ fragmentation mechanism and dust formation, as well as to estimate the amount of breakage and dust that could be transported out of the bed. For this purpose, a coupling between the open-source DEM (Discrete Element Method) code LIGGGHTS and the open-source CFD (Computational fluid dynamics) code OpenFOAM is being used in this work. First, dust transport investigations are discussed, focusing on the transport of a group of particles as a function of their size. Here, large particles show the tendency to spread more in the pebble bed and seem to be subject to a non-negligible drag to gravitational force ratio. Particles with diameters below 20 microns exhibit a higher clogging probability, which then remains relatively constant until the particle’s size gets close to the average pore size of the pebble bed. When particles reach about 110 microns in diameter, the clogging probability sharply increases to nearly 100% within a travel distance of 40 mm.
Afterwards, the plausibility of previously carried out thermo-mechanical FEM analyses on the HCPB concept of DEMO will be discussed. This will be done by comparison to DEM simulations where a sensitivity DEM study on a portion of the pebble bed will be presented. In particular, stresses rising from the thermal expansion of the ACB pebbles and related fracture probability when varying parameters such as initial PF, pebble size distribution, and temperature amplitude will be discussed.

Speaker: Dario Passafiume

31 Molecular dynamics simulations of helium diffusivity in lithium orthosilicate crystals using a machine learning potential

Ruijie Zhang, Takuji Oda Department of Nuclear Engineering, Seoul National University

Abstract:
Li₄SiO₄ is a promising solid tritium breeding material due to its high lithium density, thermal stability, and excellent tritium release performance. The Li-n reaction inevitably produces helium atoms, which are initially produced in bulk. Helium atoms have a closed-shell structure and small atomic radius, resulting in low solubility and a tendency to be trapped in defects such as dislocations, grain boundaries, and especially vacancies. Their accumulation causes crystal swelling and degrades mechanical properties. Hindering the diffusion of He atoms can prevent the coalescence and growth of He bubbles. Therefore, understanding the microscopic behavior of He atoms in crystals is essential.
Atomistic simulations such as density functional theory (DFT) calculations and molecular dynamics (MD) calculations help to uncover diffusion mechanisms and accumulate diffusion data. In this study, to realize accurate MD simulations, we construct a machine learning potential using DFT calculation data as the training set. The moment tensor potential (MTP) is adopted.
We study the He diffusivity in two cases: (1) the diffusion of a He atom in the perfect crystal as an interstitial atom (Dint); (2) the diffusion of a He atom interacting with a Li vacancy (DLi). At low temperatures (< 800 K), Dint exhibits anisotropy, with faster diffusion along the y-direction, while DLi remains nearly immobile due to strong He -Li vacancy binding. At high temperatures (>800 K), Dint converges to DLi because Li atoms start to diffuse significantly, and He is no longer deeply trapped in a Li vacancy.

Speaker: ruijie zhang (Department of Nuclear Engineering, Seoul National University, Seoul 08826.)

32 Optimized Design of HCCB Breeding Blanket with Casing Structure for Enhanced TBR in DEMO Reactor

Xiaoyong Wang, Ziqiang Zhao, Zaixin Li
Southwestern Institute of Physics, No.5 Huangjin Road, Shuangliu District, Chengdu City, Sichuan Province, China

The breeding blanket is a key component of the DEMO reactor, essential for tritium breeding, energy extraction, and nuclear shielding. To achieve a higher Tritium Breeding Ratio (TBR), a novel helium-cooled ceramic breeder (HCCB) blanket with a casing structure has been proposed. This design features two concentric pipes: the inner pipe contains lithium silicate, and the annular gap serves as a helium cooling channel. Surrounding the outer pipe is a beryllium (Be) pebble bed, which enhances the blanket's performance.
The casing structure is designed for simplicity and robustness. Using two concentric pipes simplifies the accommodation of lithium silicate and the helium cooling channel, ensuring structural integrity and ease of manufacturing. This reduces the number of welds, minimizing potential weak points and enhancing reliability. The structure also optimizes material usage, reducing costs while maintaining high performance.
A comprehensive optimization analysis covering neutronics, thermo-hydraulics, and mechanics has been conducted. Neutronic analysis shows a TBR of approximately 1.2 for the entire DEMO reactor, even with windows for diagnostic and heating systems. Thermal-hydraulic results indicate that the temperatures of the Be pebble bed (561℃), lithium orthosilicate (870℃), and reduced-activation ferritic/martensitic steel (RAFMs) (538℃) are all below their respective limits, ensuring thermal stability and safety. Thermal-mechanical analysis shows a maximum stress of 265 MPa during normal operation and 371 MPa during an in-box Loss of Coolant Accident (LOCA), meeting safety requirements.
The HCCB breeding blanket with a casing structure represents a significant advancement in DEMO reactor technology. Its simple and robust design, reduced welds, and optimized material usage enhance tritium breeding efficiency and reactor safety. Future work will focus on refining the design, conducting experimental validations, and exploring scalability for practical application. This design provides valuable insights for future breeding blanket technologies.

Speaker: Xiaoyong Wang

33 Introduction on Tritium Transport Analysis Platform – THETA-FR to Predict Tritium Behaviour in Tritium Breeding Blanket System

Yonghee Lee1, Alice Ying2, Mu-Young Ahn1, Hyung Gon Jin3, Sungbo Moon1, Myungho Kim1
1Korea Institute of Fusion Energy, Daejeon, Korea
2University of California, Los Angeles (UCLA), LA, USA
3Korea Atomic Energy Research Institute, Daejeon, Korea

To implement a tritium breeding blanket in a fusion reactor, extensive research and efforts are required. Among these, the development of a tritium transport analysis model is critical for establishing both the safety and design criteria of the tritium breeding blanket. While the tritium is transported from the breeding blanket to the tritium facility, some of tritium may be released into external environment, which can be a serious risk to the working environment. Accordingly, to ensure operational safety of the breeding blanket system, the Korea Institute of Fusion Energy (KFE) has started to develop the tritium transport analysis platform, THETA-FR (Tritium/Hydrogen Enhanced dynamic Transport Analysis platform for Fusion Reactor), in collaboration with UCLA. THETA-FR is based on COMSOL Multiphysics and MATLAB Simulink, and the development of the components and the integrated system of THETA-FR has been completed recently. THETA-FR simultaneously performs dynamic heat transfer analysis and H isotope diffusion analysis at the sub-system level for each component of the breeding blanket system and integrates the results at the system level to expect the comprehensive tritium behavior. Through THETA-FR, it is expected that the effective design guidelines for tritium management and safety in the breeding blanket system can be established.

Speaker: Yonghee Lee

10:30 AM Coffee break

Thermal properties

34 Pebble Beds Thermal Expansion Measurement with Radial Expansion Correction of the Sample Holder

Harsh Patel1, *, Maulik Panchal1, Aroh Shrivastava1, 2, Paritosh Chaudhuri1, 2
1Institute for Plasma Research, Bhat, Gandhinagar – 382428, India
2Homi Bhabha National Institute, Anushaktinagar, Mumbai – 400094, India

An experimental setup was developed to measure the coefficient of thermal expansion (CTE) of lithium ceramic pebble beds within the typical operating temperature range of fusion reactor blankets. The setup enables accurate characterization of the thermal expansion behaviour of granular ceramic materials, which is critical for understanding the thermo-mechanical performance of breeder zones in fusion blankets.
The pebble bed is formed by filling lithium metatitanate (Li2TiO3) pebbles into a cylindrical container, compacted slightly and confined by a cylindrical block placed on top. Axial expansion of the bed is captured using a Linear Variable Differential Transformer (LVDT), which is mechanically connected to the container and the top block via two precision quartz rods. These rods act as displacement transmitters, allowing accurate transfer of bed expansion to the LVDT while thermally isolating it from the high-temperature zone. To provide uniform heating across the bed height, a three-zone split-tube radiation furnace surrounds the assembly. The heating rate is kept sufficiently low to ensure that the entire pebble bed attains thermal equilibrium during each measurement interval, thus avoiding transient temperature gradients within the bed that could affect measurement accuracy.
Correction mechanisms were implemented to isolate the thermal expansion of the pebble bed from that of the setup components. Two separate correction curves were developed. The first correction was obtained experimentally by replacing the pebble bed with a solid cylindrical quartz block of the same height and measuring the system’s axial expansion using the same method. The intrinsic expansion of the quartz block, measured using a calibrated dilatometer, was then added to this axial expansion to generate an accurate correction curve. The second correction accounted for the radial expansion of the container, which could alter the contact and packing conditions of the pebbles. This was addressed analytically and through Finite Element Method (FEM) simulations, allowing the derivation of a precise radial expansion compensation factor.
The setup was validated by measuring the thermal expansion of reference materials using cylindrical blocks of the same height as the pebble beds, and the results were found to be in good agreement with standard values. The experimental setup, calibration methodology, and representative results will be presented and discussed in detail.

Speaker: Harsh Patel

35 Thermal Cycling Test of ACB Pebbles in Comparison with Standard Long‐Term Annealed Samples

Julia Leys, Thomas Bergfeldt, Oliver Leys, Regina Knitter
Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM), Karlsruhe, Germany

Advanced Ceramic Breeder (ACB) pebbles represent the EU reference material for solid tritium breeders. Their application is foreseen in the ITER TBM and in DEMO. Therefore, comprehensive knowledge about their material properties are essential for the development of blanket concepts such as the Helium Cooled Pebble Bed (HCPB) or the Water cooled Lead Ceramic Breeder (WLCB) designs. The ACB material has been extensively evaluated in the past, also with regard to their long-term thermal stability, but this study presents results from the first thermal cycling test. Even if the pebbles will not experience such a thermal cycling in a future fusion power plant, the experiment was used to test the material under extreme thermal stress.
For the thermal cycling test, two batches of ACB pebbles consisting of nominal 70 mol% Li4SiO4 and 30 mol% Li2TiO3 were used. The samples experienced cycles with 12 h at 300 °C and 12 h at 800 °C. The ACB pebbles were retrieved after 2, 14 and 30 cycles. A gas atmosphere of He + 0.1 vol% H2 was applied during the experiment. The ACB pebbles underwent a comprehensive characterisation using ICP-OES, XRD, SEM, uniaxial crush load tests, He-pycnometry, Hg-porosimetry before and after the heat treatment.
The results of the first thermal cycling test on ACB pebbles will be presented. Furthermore, the outcomes will be compared to previous long-term annealing experiments. This study will contribute to the knowledge on thermal stability of these ceramic breeder pebbles.

Speaker: Julia Leys (Karlsruhe Institute of Technology)

36 Effect of Li-depletion on thermal conductivity in Li2BeSiO4

Daigo Kanamori, Keisuke Mukai, Makoto. I. Kobayashi,
a The Graduate University for Advanced Studies, SOKENDAI, Toki, Gifu 509-5292,Japan
b National Institute for Fusion Science, Toki, Gifu 509-5292, Japan
Keywords : Solid breeder materials, Li-depletion, thermal conductivity, molecular dynamics

To realize fusion reactor, it is necessary to develop tritium breeders and neutron multiplier materials. We focus on Li-Be oxides which can work as both breeder and multiplier. Especially, Li2BeSiO4 is possible to work as an accident tolerant material because of its extremely low reactivity with water vaper [1]. Also, Samolyuk [2] reported that Li2BeSiO4 have thermal conductivity about three times higher than that of the conventional candidates. Since the higher thermal conductivity is an advantage for a fusion reactor, Li2BeSiO4 can be the promising candidate for fusion blanket. However, it is foreseen that Li concentration would be decreased due to burn up under a reactor operation. The loss of Li introduces vacancies at Li sites, leading to reduction of thermal conductivity. It is necessary to evaluate the effect of Li-depletion on the thermal conductivity for the application of Li2BeSiO4.
In this study, we investigated the effect of Li-depletion on thermal conductivity of Li2BeSiO4 using experiments and molecular dynamics simulation. LiOH·H2O, BeO and SiO2 will be mixed and sintered in an atmosphere controllable furnace to synthesize Li2xBeSiO4 (x=1, 0.95, 0.9, and 0.8). The structure of these synthesized specimens will be analyzed by X-ray diffraction (XRD) to estimate the phase compositions such as Li2BeSiO4, BeO and Be2SiO4 . The thermal diffusivity are to be measured by laser-flush method. Furthermore, calculation results of thermal conductivity using Green-Kubo methods with universal machine learning potential will be reported at the workshop. The effect of Li-depletion on thermal conductivity will be discussed from these results.

Speaker: Daigo Kanamori

37 Preliminary Engineering Analysis for CN HCCB TBM Regarding ITER New Baseline Scenario

Xinghua Wu, Shen Qu, Ruyan Li, Qixiang Cao, Fengchao Zhao, Long Zhang, Xiaoyu Wang
Southwestern Institute of Physics,P.O.Box 432, Chengdu, 610041, China

Among the different TBM concepts proposed for DEMO design of different countries, China finally determined to develop the Helium Coolant Ceramic Breeder (HCCB) TBM, following the strategy of national magnetic confined fusion energy development. The conceptual design of HCCB TBM has been completed since 2015, which was composed of four sub-modules with back plate, and each sub-module consist of the first wall (FW), the cap, the rib, the breeding zone with Li4SiO4 and Be pebbles, and the back manifold zone, cooled by 8 MPa Helium.
During preliminary design phase, some design update for CN HCCB TBM has been carried out, considering engineering performance and manufacturing feasibility. The previous version of the ITER Research Plan (IRP) was developed following 2016 staged approach baseline, which foresee first plasma in December 2025 and fusion power operation from 2035 to 2041. However, considering some engineering and technical issues, i.e. challenge with regard to licensing, accumulated delays in assembly of tokamak and repairs needed on vacuum vessel (VV) and thermal shield (TS), replacement of Beryllium with Tungsten on first wall (FW), a new ITER Baseline has been under development since the beginning of February 2023, which addressed a new ITER operation strategy allowing to start the nuclear phase as soon as possible.
Regarding ITER 2024 new baseline scenario, further design optimization and transient thermo-hydraulic analysis was performed for CN HCCB TBM, and the results showed that for the 2024 baseline, the maximum temperature of structural and functional materials have been largely reduced, especially for the tritium breeder Li4SiO4 pebble bed, which temperature has been reduced for 20.6%; through adding electrical heater inside tritium breeder zone, the maximum temperature of Li4SiO4 could increase to 821℃, and the average temperature was increased to 604℃, which can basically meet the requirement of tritium release temperature. According to calculated temperature distribution for three operational conditions, system-level transient tritium transport analysis was further performed, the results showed that adding electrical heater inside tritium breeder zone was enough for tritium balance requirement, it's not needed to further add electrical heater inside neutron multiplier zone. Since there was no major change for overall temperature distribution of CN HCCB TBM after adjusting the bypass flow rate and adding internal electrical heater, especially for the FW region and breeding zone, the structural performance was generally the same with previous baseline, only with slight difference on TBM back plate and pipes inside TBM shield. Preliminary structural analysis results showed that the primary and secondary stresses were well below the limit of structural material, which could ensure the overall structural integrity.

Speaker: Xinghua Wu

38 Dual-Direction Hot-Wire Measurements of Anisotropic effective thermal conductivity in Fusion Pebble Beds

Maulik Panchal1,, Harsh Patel1, Paritosh Chaudhuri1,2
1Institute for Plasma Research, Bhat, Gandhinagar - 382428, India.
2Homi Bhabha National Institute, Anushaktinagar, Mumbai - 400094, India.
E-mail address: maulikpanchal@ipr.res.in
Accurate prediction of thermal performance within the breeder zones of fusion blankets is critically dependent on the effective thermal conductivity (keff) of ceramic pebble beds. These beds, typically composed of lithium-based ceramics such as lithium titanate (Li₂TiO₃), exhibit anisotropic thermal behaviour when subjected to significant mechanical stresses and steep thermal gradients during operation of fusion reactor. While numerous studies have reported keff values under isotropic assumptions, such simplifications overlook the inherent directionality in heat transport arising from random packing, surface roughness, preferential contact orientation, and the non-spherical shape of pebbles. Previous studies have often assumed isotropic keff, which can lead to significant inaccuracies in thermal modeling, potentially impacting tritium breeding efficiency and structural integrity.
To address this gap, we have developed and implemented an experimental methodology based on the transient hot-wire technique to independently measure keff in both axial and radial directions. The experimental setup incorporates dual hot-wire probes aligned along orthogonal axes, enabling direct quantification of anisotropic keff in compressed pebble beds. This approach allows for systematic evaluation of how mechanical compaction influences directional heat transport in realistic breeder configurations. The experiments are designed to be conducted across a range of compressive stresses and at temperatures relevant to fusion blanket operation. The experimental methodology, measurements, and key findings will be presented and discussed.

Speaker: Maulik Panchal

12:40 PM Closing

12:45 PM Lunch break

1:45 PM Meeting and walking for technical tour

Meeting time 13:45
Meeting point: Kihada hall
Walking takes about 10 min

Technical tour

Technical tour at Heliotron J plasma facility at Kyoto University Uji campus will be held on 8th October from 14:00 (about 60 min).

What is Heliotron J?
Heliotron J is an experimental plasma device at Kyoto University designed to study and optimize the helical-axis heliotron configuration for magnetic confinement fusion. It features advanced magnetic field control to improve plasma confinement and stability, aiming to advance fusion reactor research by exploring new helical magnetic field concepts and enhancing particle confinement
http://iae.kyoto-u.ac.jp/heliotronj/en/index.html

IEA subtask meeting