Ninth DEMO and Fusion Plants Workshop

10 Jun 2025, 09:00 → 13 Jun 2025, 19:05 Europe/Vienna

Aomori, Japan

Key Deadlines

30 April 2025	Deadline for submission of abstracts through IAEA-INDICO for unsolicited posters.
30 April 2025	Deadline for submission of application for participation via the InTouch+ platform
15 May 2025	Notification of acceptance of abstracts and of assigned awards
10 June 2025	Event begins
13 June 2025	Event ends

 

Although ITER is a multinational project, there are no such plans for DEMO and fusion plants. Rather, these concepts are being developed by individual governments, private companies and some public–private joint ventures. Such a diverse development framework can enable fast progress and increased innovation, but it is important to maintain strong international cooperation and government support to ensure that fusion energy can come to fruition. Against this backdrop, the IAEA has established a series of Workshops to facilitate international collaboration on defining and coordinating DEMO and fusion plants development activities.

Topics

The topics for the Ninth edition of the DEMO workshop cover:

Topic 1: Magnets: 

The role and industrial realization of high temperature superconductors for DEMO and fusion plants, including implications for performance, timelines, and test facilities.

Topic 2: Tritium Fuel Cycle

Readiness of sub-systems required for DEMO and fusion plants, addressing tritium inventory concerns (e.g., tracking, mobility, and monitoring), initial supply in a competitive market, and the international landscape of research efforts.

Topic 3: Neutronics

Considerations for magnets, tritium breeding performance, volumetric neutron sources and neutronics tool development.

 

Information Sheet.pdf

Venue Travel for DPW-9.pdf

Tuesday 10 June

09:10 → 09:30 Opening

09:30 → 10:55 Magnets

09:30 Introduction - Day 1

Speaker: Nagato Yanagi (National Institute for Fusion Science)

09:40 Development of LTS and HTS wires

10:15 HTS Magnets Development for the UK’s STEP Programme

The Spherical Tokamak for Energy Production (STEP) is an ambitious public programme to deliver a UK prototype fusion power plant alongside a pathway to commercial deployment. One of the primary technical challenges facing the programme is the development of the unprecedented scale high temperature superconducting (HTS) magnets required by the reactor concept design.

This presentation outlines the key design features of these magnets following the latest iteration of the STEP concept. These include the remountable toroidal field (TF) coils, which enable the power plant’s vertical maintenance strategy, and the replaceable ‘central magnet unit’, which comprises the inner limbs of the TF coils, the inner divertor shaping coils, and the central solenoid. The significant design integration, technology, and manufacturing challenges are summarised.

To address these challenges, The STEP Magnets Technology Development Programme planned for the period 2025 – 2029 is presented, culminating in the delivery of a sub-scale toroidal field model coil demonstrating the key technologies required and providing confidence in the scale-up to production. Developing HTS capability, from modelling to large-scale manufacturing and testing, is central to successful delivery of STEP and eventual commercialisation. This presentation outlines the public-private partnership model UK Industrial Fusion Solutions Ltd (UKIFS) is implementing to realise this ambitious programme.

Acknowledgement:
This work has been funded by STEP, a major technology and infrastructure programme led by UK Industrial Fusion Solutions Ltd, which aims to deliver the UK’s prototype fusion powerplant and a path to the commercial viability of fusion.

Speaker: Ezzat Nasr

10:55 → 11:10 Coffee Break

11:10 → 12:30 Magnets

11:10 Magnet development for national project 2 (JA-DEMO)

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

11:50 Recent Challenges on REBCO Conductor and Magnet for High Field Applications and Fusion

A notable progress has been made over the last decades in REBCO conductor and magnet technologies. Now multiple companies routinely deliver commercial-level REBCO conductors in different recipes, while various REBCO “user” magnets are currently under construction and some are even in routine service, though at low magnetic fields mostly less than 20 T. Despite the outstanding achievements by our community in both conductor and magnet, we are still struggling with critical technical challenges that limit wide spread use of REBCO beyond laboratory magnets toward general industrial use. Spatial and temporal distributions of currents—transport, screening and radial leak in case of no-insulation—are still unclear, thus precise estimation of critical current and peak magnetic stress is still challenging. As a result, modern REBCO magnets have been designed and operated without knowing their electrical and mechanical limits; to date REBCO magnets that “repeatedly” reach 20 T or greater are rare and none are under routine service. This talk begins with introduction to the recent progress in REBCO conductor and magnets, summarizes key observations and challenges, and discuss potential approaches toward practical application of REBCO technology for high field applications, not limited to fusion.

KEYWORDS
Challenge, fusion, high field, magnet, REBCO

Acknowledgement
This work was supported in part by National R&D Program through the National Research Foundation of Korea(NRF) funded by Ministry of Science and ICT(2022M3I9A1073924), in part by National R&D Program through the National Research Foundation of Korea(NRF) funded by Ministry of Science and ICT(2022M3I9A1072846), and in part by the Applied Superconductivity Center, Electric Power Research Institute of Seoul National University.

Speaker: Seungyong Hahn

REBCO.for.HighFieldMagnet_20250518_SH.docx

12:30 → 13:50 Lunch Break

13:50 → 15:10 Magnets

13:50 Test facility 1 (CFTER/CRAFT/BEST)

14:30 SUCCEX, 16 Tesla Superconducting Conductor Test Facility

Korean Fusion Energy Development Promotion Law (FEDPL) was enacted in 2007 to promote a long-term cooperative fusion research and development among participating industries, universities and research institutes. As a following step, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) has been initiated in 2012. As a result of a conceptual design study for K-DEMO, the major and minor radii are 6.8 and 2.1 m, respectively, and toroidal field magnets of K-DEMO can generate around 8 T at plasma center with a peak magnetic field of ∼16 T. For conductor tests under such a high magnetic field, a new conductor test facility is required and a 16 T conductor test facility is under construction. The conductor test facility, named SUCCEX (SUperConducting Conductor EXperiment), will feed the sample with a current up to 100 kA and the sample temperature will be varied from 4.5 K to ∼20 K. The 16 T Nb3Sn magnet system, with a bore of 0.6 m diameter, is divided into two concentric split pairs, inner coil (IC, peak field of 16.2 T) and outer coil (OC, peak field of 12.4 T), connected in a series, with a nominal operating current of ∼24.8 kA. The overall design and the current status of the project will be presented.

Speaker: Keeman Kim (Korea Institute of Energy Technology)

15:10 → 15:25 Coffee Break

15:25 → 16:30 Magnets

15:25 Neutron Irradiation Effect on Superconductivity of ReBCO Tapes

Deuterium and tritium reaction generates 14 MeV neutrons and most of the generated neutrons are captured by blankets and plasma vacuum vessel (VV) and the energy of these neutrons will be exchanged to electricity. The rest can penetrate the components and reach superconducting (SC) magnets outside of the VV. A recent study on neutron mapping in fusion reactors has expanded to include thermal neutrons, and one research shows the huge number of thermal neutrons would exist in the SC magnets. At the same time, ReBCO tapes have been developed. Therefore, GdBCO, EuBCO and YBCO tapes were taken as test materials for the research on neutron irradiation effect on the superconductivity. The neutron irradiation was carried out at Japan Research Reactor #3 (JRR-3) located at Tokai in Japan. The maximum fast and thermal neutron fluence were 1.46 x 1021 n/m2 and 8.29 x 1022 n/m2, respectively. Several noble researches have been already conducted at Vienna using TRIGA MARK II. Since JRR-3 has perfectly different neutron flux envelope from TRIGA, the effect of the thermal neutron was focused on in this study. In addition, the ReBCO layer was pealed and scratched out after the irradiation and the analysis with a Ge detector was carried out to check the isotopes in ReBCO layer.
153Gd was detected by the Ge detector after the irradiation and this is good evidence that Gd transmutation occurred during the irradiation. The GdBCO irradiated with no Cd shielding showed no superconductivity even at 5 K. The shielding with 75 μm and 125 μm thick Cd foil has significant effect in preventing degradation of the critical temperature (TC) and this is also clear evidence that the thermal neutron degrades the superconductivity of GdBCO.
Gd has five stable isotopes and 155Gd and 157Gd have huge cross sections for thermal neutron on the order of 106 barns. This {n,γ} reaction will be the reason for the degradation. There are three considerations on the mechanism for the degradation.
(1) The Gd atoms will be released by the recoil from the {n,γ} reaction.
(2) Oxygen atoms will be removed from the perovskite crystal, creating lack of oxygen.
(3) Exchange of electron during Gd transmutation will disturb the electric potential on the CuO2 planes.
It is expected that the further study will make the mechanism clear.
On the other hand, EuBCO without Cd shielding showed some degradation but the Cd shielded samples presented the same properties as the non-irradiated sample. YBCO does not show the degradation at all by the irradiation, and TC was improved a little by the fast neutron. This improvement would be caused by the relaxation of internal strain in the YBCO layer.
The information on IC and BC2 after the irradiation will be presented at the workshop.

Speaker: Arata Nishimura (Japan)

Abstract to WS AOMORI r1 20250506.pdf

16:05 Discussion

16:25 Summary of the session – Status and prospects of fusion magnet development in the world –

A summary of the session on the development of fusion magnets in the world is given from the seven talks and the overall discussion. The session covers the topics on “Development of LTS and HTS wires”, “Magnet development for national projects”, “Magnet development for public-private partnership”, “Test facilities”, and “Irradiation”. Highlights from the talks are reviewed, and the future prospects are discussed.

Speaker: Nagato Yanagi (National Institute for Fusion Science)

Yanagi IAEA demo programme workshop 2025 abstract.pdf

16:30 → 16:45 Coffee Break

16:45 → 17:55 Special Topic

16:45 DT - Spherical Tokamaks

Speaker: Howard Wilson (UK Industrial Fusion Solutions)

17:15 National Strategies of China (CNDA) - TBC

17:25 National Strategies of Europe

Speaker: Francesco Maviglia (EUROfusion, PPPT Department, Building R3 Boltzmannstr. 2 Garching 85748, Germany)

17:35 National Strategies of India (TBD)

17:45 National strategies of Japan - Fusion research and development strategy for JA DEMO

Japanese fusion program has three stages for realization of fusion energy, i.e. establishment of physics/engineering bases, fusion energy production and electricity production. We are in the stage of fusion energy production, where QST is contributing to the ITER project for the demonstration of 500 MW fusion energy production with Q=10. QST is also implementing the Broader Approach (BA) Activities in EU-JA collaboration for support and supplement of the ITER project. QST aims at early transition to the next stage of electricity production by integrating the ITER project, the BA activities and domestic activities. In the next stage, construction of JA DEMO is planned, where electric power of >100 MW will be generated.
Recently, research and development activities for the realization of fusion energy are globally being accelerated both in public and private sectors from the perspective of the transition towards a Net-Zero Society. In Japan, “Fusion Energy Innovation Strategy” was formulated in Cabinet Office as Japan’s first national strategy for fusion energy in April 2023. This strategy presents the vision of “Commercialization of fusion energy” and highlights development of the fusion industry as well as development of fusion technology. In order to promote the strategy, it is required that establishment of framework for conducting R&D by bringing together, centering on QST, academia and private companies (fusion technology innovation hub). Furthermore, “Integrated Innovation Strategy 2024 (Cabinet decision)” stated that “Japan will aim to realize fusion energy as soon as possible by preparing a timetable that includes necessary national efforts toward achieving the first demonstration of power generation in the 2030s ahead of other countries”. Considering the present situation, the revision of Fusion Energy Innovation Strategy is planned this spring.
In accordance with the recent situation described above, QST is investigating phased approach strategy to accelerate JA DEMO program with the same TFC size as ITER. The objective for each phase is demonstration of electricity production with almost zero net electric power in Phase I; demonstration of tritium breeding with breeding blankets in Phase II; and demonstration of steady-state operation with 100 MW level net electric power using high  and high confinement plasma as well as improved breeding blankets and heating system in Phase III. As phase changes, enhanced plasma performance, improved blankets and high efficiency heating system are required. In order to utilize key technologies acquired through the ITER project and the BA activities for the acceleration of the DEMO project, QST has proposed to enhance facilities and equipment in Rokkasho and Naka Institutes such as facilities of fuel cycle, blanket, neutron source, superconducting magnet, plasma heating.

Speaker: H Takenaga

DEMO-WS_Takenaga.docx

17:55 → 18:00 Closing Session for the Day

Wednesday 11 June

09:30 → 10:50 Tritium Fuel Cycle

09:30 Chair Introduction

Speaker: Rachel Lawless (UKAEA)

09:40 Introduction and Topic overview summary

Outline the aims of the session, give a broad overview of TFC developments to set up deeper talks without them covering all details

Speaker: Christian Day (Karlsruhe Institute of Technology)

10:10 Fusion for Energy activities on the ITER fuel cycle and the European Roadmap for fuel cycle technology developments

Fusion For Energy is responsible for delivering large units of the ITER fuel cycle. This involves extensive design and technology developments for the tritium plants’ Isotope Separation and Water Detritiation systems in preparation for manufacturing. The manufacturing of the cryogenic adsorption pumps for the ITER torus and plasma pumping is well advanced, and the delivery of the fully tritium-compatible Torus Cryopumping System to ITER is nearly complete.
In parallel with the procurement activities for ITER, Fusion for Energy launched a Technology Development Programme (TDP) in 2024 as part of its Industrial Policy implementation actions. This TDP aims to build and reinforce European Fusion Supply chain capabilities for critical future commercial fusion technologies.
To define the fuel cycle roadmap, it was first necessary to identify the fuel cycle key technologies and their current technology readiness level. Fusion for Energy organized a Fuel Cycle workshop that covered vacuum pumping, storage and injection, fuel purification, isotope separation, water detritiation, air detritiation and tritium management. Starting with an online event in February 2025 an exhaustive list of fuel cycle technologies was prepared to complete the drafted fuel cycle technology map.
During the in-person workshop in March 2025, Fusion for Energy brought together academia, research laboratories, industry, start-ups and the ITER Organization to discuss and characterize each of these technologies. Experts from all relevant technology fields contributed to creating a database with the current readiness level of the mapped technologies. The database includes information on the applicability of the technologies, available test facilities, active and interested European entities, and the necessary next steps to advance technology’s maturity. The 80 participants developed a comprehensive technology database that allows for defining a European roadmap for the Fusion Fuel Cycle domain.
The outcome will serve all stakeholders to guide their actions in their respective domains and interests, allowing an effective investment of resources. Given the fast evolution of technology, a periodical follow-up of the workshop outcome is assured in subsequent technology mapping exercises.

The presentation will provide a brief overview of the current F4E activities for the ITER fuel cycle procurements and will outline the new activities of the Fuel Cycle Technology Development Programme, including the 2025 European Roadmap for fusion fuel cycle developments.

Speaker: Matthias Dremel (Fusion for Energy)

Topic1_Dremel_Abs.docx

10:50 → 11:10 Coffee Break

11:10 → 12:30 Tritium Fuel Cycle

11:10 An overview of tritium inventory management progress and challenges, and the role of the UKAEA – Eni H3AT facility in supporting the fusion community

Tritium inventory management is a critical challenge for the fusion community.
Firstly, it is important to understand the quantities of tritium required to operate a fusion power plant, along with its distribution within the fuel cycle, and its physical and chemical forms. This knowledge will support effective and proportionate regulation of fusion facilities. Moreover, the necessary improvements to predictions of tritium’s behavior in systems and materials supports the optimization of system requirements and sizing.
Secondly, accurately assessing tritium quantities across systems facilitates minimization of tritium inventory in fusion facilities. This is essential for reducing cost, enhancing safety and facilitating the development and implementation of regulations, as well as ensuring that start up inventory demand does not exceed supply.
Whilst critical, understanding inventory requirements is a significant technical challenge. This talk will explore the nature of this challenge and review the range of estimates found in the literature, commenting on assessment methods employed. The approach to this problem taken by UKAEA will also be explored.
In addition to understanding inventory requirements in advance of deployment, fusion facilities will also need to track tritium migration through various systems once they are operational. There are four main reasons why this is necessary:
• Process control
• Safety
• Environmental protection
• Non-proliferation
Each of these areas will have differing requirements in terms of measurement accuracy, uncertainty and frequency. Additionally, safety, environmental protection and non-proliferation all fall under regulatory oversight. Given the immaturity of regulatory environments for fusion, defining precise analytical requirements for tracking tritium remains challenging.
Further complicating this issue, many of the diverse environments within the fusion fuel cycle require specialized measurement techniques, necessitating the development and validation of multiple technologies. Current understanding in this area will be presented along with suggestions for further work to address knowledge and technology gaps.
In order to address key challenges in fusion fuel cycle development, UKAEA is constructing the UKAEA-Eni H3AT Tritium Loop facility at its Culham site in partnership with Eni and supported by collaborations and contracts with ITER Organization and AtkinsRealis, respectively. The 100g tritium inventory pilot plant scale facility will for the first time demonstrate a closed loop, continuous flow system, enabling testing of subsystem technologies, including dynamic responses, as well as validation of UKAEA developed fuel cycle models. Additionally, the facility will include substantial experimental capacity for off-loop tritium research. An update will be provided on the facility’s progress and its anticipated benefits for the fusion community will be outlined, including in addressing the challenges of inventory management.

Speaker: Rachel Lawless (UKAEA)

Topic2_Lawless_Abs.docx

11:50 Canadian and International Fusion Fuel Cycle Capabilities supporting Global Fusion Energy Developments

Since the dawn of the CANDU nuclear power technology, its significance as a tritium fuel supply source for fusion developments has been well recognized. The tritium by-product from the capture of neutrons by the heavy water in CANDU reactors is removed for safe operation of the reactors, stored in solid form and supplied for various peaceful applications including fusion energy research developments. For example, the experiments in the JET facility were carried out with tritium purchased from Ontario Power Generation’s (OPG) Darlington Tritium Removal Facility (DTRF) and packaged and shipped by Canadian Nuclear Laboratories (CNL). Many fusion developers are potentially reliant on the availability of tritium from continued operation of the CANDU reactors around the world. Recent estimates of the amount of tritium potentially available from CANDU reactors (globally) is about 30 to 40 kg over the next three decades; however, with ITER operation potentially requiring tritium in the 10 kg range in late 2030’s, the tritium inventory is expected to dwindle there on. This potential demand for ITER can cause a critical tritium supply limitation for fusion development demonstrations around the world. Canada’s ambition to deploy CANDU® MONARKTM 1,000 MWe reactors may partially relieve the tritium supply limitation in the second half of this century.
CNL’s Tritium Facility is a one-of-a-kind laboratory licensed to handle up to 1MCi for processing and up to 2.5MCi in storage. This facility, under a contract with OPG, dispenses tritium for OPG customers around the world. And, in conjunction with the Hydrogen Laboratory at CNL where other isotopes of hydrogen are handled, this facility has been the Canadian centre for establishing a range of tritium capabilities. Some examples are: hydrogen-water isotope exchange processes for heavy water production and detritiation of heavy and light water for tritium management/control; air detritiation; materials development for tritium services in high temperature fission and fusion reactor applications; tritium permeation studies in metals and membranes; tritium pumping; tritium analytics, diagnostics and accountancy; tritium storage, packaging and transportation to national and international customers; specialized glove boxes to handle tritium processing operations; and ventilation systems for safe operation in closed spaces. These capabilities have been demonstrated in laboratory operations and in small- and large-scale systems over the last several decades.
The recent boom in new fusion companies has put these capabilities in the spotlight for the benefit of fusion energy across the world. The UNITY-2 facility, currently under detailed design at CNL, encompasses Canada’s tritium capabilities and Kyoto Fusioneering’s engineering capabilities to deliver a versatile D-T fusion fuel cycle platform with up to 30 g tritium inventory for fusion energy developers to collaborate, witness, study and validate technologies, as well as test out their proprietary equipment and processes. The UNITY-2 facility will consist of all processes and components typically required for the fusion fuel cycle in a power plant, but at prototypical conditions.
This presentation will discuss current status of tritium supply, Canadian fusion capabilities and their applications in UNITY-2 for advancing fusion energy developments.

Speaker: Sam Suppiah (Canadian Nuclear Laboratories)

Topic3_Suppiah_Abs.docx

12:30 → 14:00 Lunch Break

14:00 → 15:20 Tritium Fuel Cycle

14:00 Progress of Tritium Plant Technologies for Tokamak Like Fusion Energy in China

14:40 Toward Technical Readiness: Private-Sector Pathways for Fusion Fuel Cycle and Power Integration

The landscape of fusion energy development is undergoing a fundamental transformation, increasingly driven by private sector initiatives. While both public and private entities target First-Of-A-Kind fusion energy systems such as DEMOs and Pilot Plants, their technical approaches and timelines differ significantly. Private sector actors emphasize accelerated deployment, productization, and market integration, often prioritizing rapid development cycles over long-term public-sector schedules. As a result, industrialization—including supply chain development and technology integration to achieve commercial-level Technical Readiness—is becoming a core focus of fusion system design.
One of the most striking changes is the early engagement of private companies in deuterium-tritium (DT) burning experiments and tritium system development, well ahead of ITER’s planned tritium operations in the late 2030s. Several private projects are expected to handle and burn tritium at significant levels from before 2030 through 2040, with the aim of demonstrating key fusion nuclear technologies such as breeding blankets, tritium extraction, and thermal energy conversion on compact platforms. These efforts mark a departure from traditional DEMO-scale programs by proposing smaller-scale, fast-track systems designed to demonstrate reliability, safety, and integration at commercially relevant scales.
This talk will overview these paradigm shifts and highlight the emerging role of public-private partnerships, focusing on the fusion tritium economy, safety protocols, regulatory frameworks, and the critical importance of societal engagement. Emphasis will be placed on the challenges of integrating nuclear and plant technologies—often identified as bottlenecks in fusion commercialization—such as materials, fuel cycles, energy conversion systems, and tritium breeding and handling.
As a case, the Japanese private-led FAST (Fusion by Advanced Superconducting Tokamak) project will be introduced. With a target start date in the mid-2030s, FAST will use a low aspect ratio tokamak with high-temperature superconducting magnets to sustain DT plasma burning for durations exceeding 1000 seconds. It will incorporate a full tritium breeding and extraction system, closed-loop thermal energy conversion, and co-generation capabilities including hydrogen production. Designed to operate at 100 MW thermal power, FAST is expected to serve as a testbed for maturing critical technologies for future fusion power plants, bridging the gap between plasma physics and energy systems engineering.
This presentation will explore how private sector innovation is reshaping fusion development timelines and technical priorities, ultimately accelerating the pathway to commercial fusion energy.

Speaker: Shutaro Takeda (Kyoto Fusioneering)

Topic5_Takeda_Abs.docx

15:20 → 15:40 Coffee Break

15:40 → 16:30 Tritium Fuel Cycle

15:40 Discussion on key points and session summary

Speakers: Christian Day (Karlsruhe Institute of Technology), Rachel Lawless (UKAEA)

16:20 Summary

Speaker: Christian Day (Karlsruhe Institute of Technology)

16:30 → 16:45 Coffee Break

16:45 → 18:05 Special Topic

16:45 Fusion Prototypic Neutron Source Risk Reduction Activity

In response to a U.S. Department of Energy (DOE), Office of Fusion Energy Sciences (FES) request for information in 2023, sixteen different concepts were submitted by the community for consideration as a fusion prototypic neutron source (FPNS). The proposed concepts vary greatly in approach, maturity, and the degree to which they accurately mimic a fusion energy system environment. To gain a better understanding of the proposed concepts, an FPNS risk reduction activity was initiated with representation from across the U.S. fusion community. The goal of the assembled team is to provide a consistent, objective, and unbiased approach to understanding and articulating the risks and benefits of different concept approaches to an FPNS. The assessment of each concept is broken into three topical areas: 1) the ability to mimic a fusion energy environment, 2) the ability to meet the performance requirements, and 3) the overall system maturity.
The approaches to estimating system maturity, performance, and ability to mimic fusion conditions will be discussed. In the case of system maturity and performance, input from the concept proposers was critical to the effort which was then expanded upon by the FPNS risk reduction team. The ability to replicate fusion conditions was assessed first for the deuteron–lithium (D-Li) stripping concept in relation to existing reference fusion concepts such as ITER and DEMO. The methodology employed to that effect was designed to bridge neutronics with irradiation-induced microstructural transformations, using a combination of neutron transport calculations, molecular dynamics (MD) simulations, chemical inventory evolution calculations, and computational thermodynamics. This methodology represents the most advanced approach to assess fusion materials evolution under irradiation to date.
The risk/benefit analysis process begins with determining the neutron spectrum and the recoil energy distributions in each material. This is followed by large scale MD simulations of high-energy displacement cascades in the range of energies dictated by the recoil distributions. Next, gaseous and solid transmutant production rates are quantified for each concept, with the resulting information being used to perform a thermodynamic analysis of emerging phases during fusion operation. We have focused on the primary fusion structural material candidates, namely silicon carbide, reduced activation ferritic martensitic steels, vanadium-based alloys, and tungsten. The impacts of these results on other international facilities using D-Li stripping sources is also considered.
The result of the FPNS Risk Reduction Activity was a report to DOE FES. This presentation will summarize the main conclusions from the report. Follow on activities have included the exploration of an integrated blanket and fuel cycle facility.

Speaker: Arnold Lumsdaine (Oak Ridge National Laboratory)

Status of the FPNS RRA - v3.docx

17:15 Overview of ITER’s Progress

17:25 National Strategies of Korea

Speaker: Keeman Kim (Korea Institute of Energy Technology)

17:35 National Strategies of Russia (TBD)

17:45 National Strategies of UK (TBD)

17:55 National Strategies of US (TBD)

18:05 → 18:10 Closing Session for the Day

Thursday 12 June

09:30 → 10:35 Neutronics

09:30 Chair Introdcution

Speaker: Saerom Kwon (National Institutes for Quantum Science and Technology)

09:40 Topic overview summary

Speaker: Lee Packer (UKAEA)

09:55 Neutronics activities in Japan

Speaker: Kentaro Ochiai (National Institutes for Quantum and Radiological Science and Technology)

10:35 → 11:05 Coffee Break

11:05 → 12:25 Neutronics

11:05 EU - DEMO/VNS

Speaker: Dieter Leichtle (KIT)

11:45 STEP (Magnets focus)

Speaker: Tim Eade (UKIFS)

12:25 → 13:55 Lunch Break

13:55 → 15:15 Neutronics

13:55 Neutronics and tritium fuel cycle R&D activities for fusion development in China

For future D-T fusion devices including DEMO reactors and plants, it is essential to achieve the high neutronics performance and to build an efficient tritium fuel cycle. There have been a few future fusion devices developed or under development in China, such as the CFETR (China Fusion Engineering Test Reactor), the burning plasma superconduting experimental tokamak device and the CFEDR (China Fusion Engineering DEMO Reactor). The neutronics calculation and analyses have been carried out to assess the tritium production capability in the corresponding breeding blankets, and the shielding adequacy to minimize the radiation impact on the superconducting magnets. Furthermore, the shutdown dose rate (SDDR) has been evaluated to address the personnel safety from the radiological dose exposure; the radioactive waste has been estimated to assess the radwaste classification. To perform the abovementioned neutornics analysis for all fusion facilities, a workflow of a series softwares and toolkits is developed and used in the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). The benchmark between the calcualtion and experimental results is also underway.
To build the D-T fuel cycle for future D-T fusion reactors, tritium inventory requirement is one of the most challenging issues for both tritium self-sufficiency and initial start-up inventory. The development of an integrated tool is underway in ASIPP. This tool is designed to simulate the process of the fuel cycle for different fusion devices, including fueling, burning, retention, purification, isotopes separation and recycling. It has been preliminarily tested and applied to design the fuel cycle for the CFEDR. In addition, a small-scale experimental facility is recently built to test the performance of the closed fuel cycle by using H/D to simulate the T fuel in order to simplify the radiation safety requirements. These tools will be used to support the design of the fuel cycle for the fusion devices newly designed in China with the aim to optimize the fuel inventory need and the efficiency.

Speaker: Shanliang Zheng (Institute of Plasma Physics, Chinese Academy of Sciences)

Abstract-20250410.docx

14:35 Contribution of the US

15:15 → 15:45 Coffee Break

15:45 → 16:25 Neutronics

15:45 Discussion on key points

Speakers: Lee Packer (UKAEA), Saerom Kwon (National Institutes for Quantum Science and Technology)

16:10 Summary

16:25 → 17:55 Special Topic

16:25 Fusion and Non-Proliferation

Speaker: Lee Packer (UKAEA)

17:55 → 18:00 Closing Session for the Day

Friday 13 June

09:30 → 12:30 Closing Session

12:30 → 14:00 Lunch Break

14:00 → 19:00 Tour of Rokkasho Institute