Choose timezone
Your profile timezone:
Ninth DEMO and Fusion Plants Workshop
→
Europe/Vienna
Aomori, Japan
Aomori, Japan
Description
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.
Contact
-
-
Opening
-
1
Opening Remarks by Mr. Takenaga, Deputy Executive Director of QST
-
2
Opening Remarks by IAEASpeakers: Mr Matteo Barbarino (International Atomic Energy Agency), Laura Wheatley (IAEA)
-
1
-
Magnets
-
3
Introduction - Day 1Speaker: Nagato Yanagi (National Institute for Fusion Science)
-
4
Current status of LTS and HTS wires for high field applications
Satoshi Awaji
High Field Laboratory for Superconducting Materials, Institute for Materials Research, Tohoku UniversityPractical Nb3Sn wires are still under improvement for high field applications such as accelerator, nuclear fusion, NMR and so on. Recently, the increase of non-Cu Jc are required for the future circular collider (FCC). In addition, the strengthening of the Nb3Sn wires have been developed. In particular, high strength Nb3Sn wires enable the compact high field superconducting magnets with a react-and-wind (R&W). We have successfully developed low temperature superconducting (LTS) magnets for the 25T- and 33T- cryogen-free superconducting magnets (CSMs) with the high strength Nb3Sn and NbTi Rutherford cables. The maximum hoop stresses are 250 MPa and 275 MP for the 14T-LTS magnets of the 25T-CSM and the 33T-CSM, respectively. In particular, the 25T-CSM has been in operation for 10 years without any problems in the LTS magnets under the high stresses.
The REBa2Cu3Oy (REBCO, RE: rare earth and Y), Bi2Sr2Ca2Cu3Oy (Bi2223) and Bi2Sr2CaCu2Oy (Bi2212) are practically available as the high temperature superconducting (HTS) wires/tapes, although the Sumitomo Electrical Industry (SEI) Co ltd. recently stopped the production of Bi2223 tapes unfortunately. The high strength Bi2223 tapes (SEI HT-Nx) were used for the 25T-CSM with the maximum hoop stress of 318 MPa. The 19T REBCO insert magnet is under construction for the 33T-CSM. The REBCO tapes have very high in-field Jc and strong mechanical tensile stress performance, although it depends on tape architectures such as thickness of Cu stabilizer and Hastelloy substrates. In addition, the artificial pinning centers (APC) such as nanorods and nanoparticles are introduced in practical REBCO tapes. It improves not only the in-field Jc but also its anisotropy. On the other hands, there are some serious issues such as delamination, slit edge, screening currents, which should be overcome.
The present status of practical LTS and HTS wires will be presented from the view point of high field applications with some recent developments of cryogen-free superconducting magnets.Speaker: Satoshi Awaji (Tohoku University) -
5
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
-
3
-
10:55
Coffee Break
-
Magnets
-
6
Magnet development for national project (JA DEMO)
Hiroyasu Utoh
National Institutes for Quantum Science and Technology,
Rokkasho, Aomori 039-3212, JapanABSTRACT
A fusion demonstration (DEMO) reactor requires toroidal field (TF) coils larger than those used in ITER and can withstand higher electromagnetic forces. This creates significant challenges regarding the manufacturability of the TF coils, the increased electromagnetic forces on the TF inner leg case, and increased fabrication costs. This means that new winding pack designs and superconducting conductors must be designed to meet these requirements, and high-strength cryogenic steel for the coil case must be developed to withstand electromagnetic forces. This paper describes these possible solutions based on the Japanese DEMO (JA DEMO) design. In the layered winding of rectangular conductors, it was found that a new conductor geometry called the hybrid R-shape conductor can significantly reduce the stress on the turn insulation while reducing the superconducting wire to less than half. In addition, a guideline for designing the conductor strand structure was provided for a superconducting conductor subjected to electromagnetic forces 1.5 times greater than those of ITER. Furthermore, the effect of composition on strength was evaluated using several prototype materials based on high-Cr austenitic steels, and the feasibility of increasing the strength of cryogenic steels was confirmed. When considering options for layer winding of rectangular conductors, we found that adopting a hybrid R-shape conductor would alleviate stress and make it possible to design an optional TF coil that could rationalize costs by optimizing the amount of Nb3Sn. The rectangular conductor layer winding method is considered a promising option for TF coils for JA DEMO, and we plan to refine the design, including manufacturability, further. Regarding conductor design compatible with high electromagnetic force and large current, it was found that it was not easy to adopt the same twist pitch as ITER for the ITER CS grade wire for the DEMO conductor. Therefore, reviewing the strand structure with a focus on the twist pitch and developing high-strength strands will be necessary. Regarding the development of high-strength structural materials, we started development with the goals of yield strength YS > 1,600 MPa and fracture toughness KIC(J): 120 MPa√m. We confirmed that our policy of increasing strength by increasing the amount of N was correct. We plan to create and evaluate materials with new compositions in the future.
Speaker: Hiroyasu Utoh (National Institutes for Quantum and Radiological Science and Technology) -
7
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, REBCOAcknowledgement
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
-
6
-
12:30
Lunch Break
-
Magnets
-
8
Cold test facilities for fusion magnet at ASIPP
Cold test is crucial to mitigate the operation risks for the fusion magnet system. In order to fit the test requirements for the next generation fusion device BEST under construction at ASIPP, a serials of SC magnet testing facilities are being built to perform the large-scale SC magnet research on mechanics, thermology and electromagnetics properties and evaluate the SC magnet system compatibility, reliability, stability, and magnet safety in fault state. Two cryostats with square section will be used for BEST TF coils tests, and up to 4 coils can be tested at once. One is with dimension of ~25m×15m×10m and the other is ~14m8m9m. Other two cryostats with circle section will be employed for BEST CS and PF coils tests with diameter of ~ 7m. The current capacity for each test system will be no less than 60 kA and each coil will be tested with operation current at 4.5 K cooled by supercritical helium. All the test facilities will be ready in 2025 based on the BEST magnets manufacture and assembly schedule.
Speaker: Huajun Liu -
9
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)
-
8
-
15:10
Coffee Break
-
Magnets
-
10
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) -
11
Discussion
-
12
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)
-
10
-
16:30
Coffee Break
-
Special Topic
-
13
SCIENCE AND TECHNOLOGY CHALLENGES IN SUPPORT OF A DT SPHERICAL TOKAMAK DEMONSTRATION FUSION PLANTSpeaker: Howard Wilson (UK Industrial Fusion Solutions)
-
14
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 -
15
National Strategies of EuropeSpeaker: Francesco Maviglia (EUROfusion, PPPT Department, Building R3 Boltzmannstr. 2 Garching 85748, Germany)
-
16
Chinese Fusion National StrategySpeaker: Prof. Minyou Ye (University of Science and Technology of China)
-
13
-
-
-
Tritium Fuel Cycle
-
17
Chair IntroductionSpeaker: Rachel Lawless (UKAEA)
-
18
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) -
19
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)
-
17
-
10:50
Coffee Break
-
Tritium Fuel Cycle
-
20
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) -
21
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)
-
20
-
12:30
Lunch Break
-
Posters
-
22
PRELIMINARY ENGINEERING ANALYSIS FOR CN HCCB TBM REGARDING ITER NEW BASELINE SCENARIO
.
Speaker: XINGHUA WU (CHINA) -
23
PROGRESS OF REBCO HIGH-FIELD FUSION MAGNET RESEARCH AT SOUTHWESTERN INSTITUTE OF PHYSICS
.
Speaker: X. Hu (Southwestern Institute of Physics) -
24
CONCEPTUAL DESIGN, SYSTEM MODELING AND OPTIMIZATION OF TRITIUM FUEL CYCLE IN A STEADY-STATE OPERATING TOKAMAK
Controlled fusion reactors with steady-state operation (SSO) will require a fundamentally new approach to the organization of the fuel cycle and the use of other innovative solutions for technological systems, since all existing and planned facilities were designed for a pulse-periodic mode in which the gas mixture is processed between discharges. Switching to long-term pulses will require stable operation of all fuel cycle systems with coordination of operating modes and fuel flows, while the total supply of tritium at the facility should be reduced as much as possible. The creation of steady-state facilities with a fuel cycle similar to the existing experimental one will lead to an unjustified accumulation of tritium.
As part of the conceptual design of controlled fusion reactors, it is necessary to estimate the amount of fuel required to start the project/facility and formulate requirements for technological systems at the technical design stage. Existing fuel cycle models for the DEMO plant predict high reserves of tritium (more than 5 kg for DEMO-FNS project with 40 MW fusion power) and do not take into account the processes inside the plasma and vacuum vessel (and oriented towards a D:T=1:1 mixture in the plasma and in all systems). A conceptual design of the fuel cycle was carried out for two FNS projects with fusion power 40 MW based on the classic JET scale tokamak and 3 MW based on a spherical tokamak FNS-ST. An effective simulation model and new codes have been developed to describe the partial flows of D and T in a vacuum vessel, as well as the processes of fuel nuclides injection, pumping in gases’ processing systems, and estimating the tritium amount in the fuel cycle systems of these plants.
A new structure is proposed of a SSO fuel cycle for FNS with "real time" processing gases without accumulating in the operational storage. It provides separate flows of fuel components and plasma isotope composition. Due to organization of several circuits (up to 13 in FNS projects), the structure provides minimal possible tritium reserves in the FNS complex. The proposed structure maintains stationary control of the plasma isotopic composition by means of a separate D/T supply using pellets (HFS, LFS - to stimulate ELMS) and heating beams. The gas puffing into a vacuum vessel without changing its isotope composition after pumping can reduce the flow rate in the hydrogen isotope separation system. This will reduce the tritium reserves required to start the facilities (less than 0.5 kg for DEMO-FNS/40 MW and less than 0.1 kg for FNS-ST/ 3 MW) by almost 10 times compared with the estimates obtained in the DEMO configuration.
As a result, a draft design of the fuel cycle of the FNS-ST and DEMO-FNS complexes was proposed and developed, being based on new technical and technological solutions. The FC composition and readiness level of tokamak technological systems realizing the D-T FNS fuel cycle have been determined. The proposed and realized approach is capable to solve the D-T fuel cycle design problems of SSO fusion devices.Speaker: sergey ananyev (nrc Kurchatov institute) -
25
Analysis of Neutron Attenuation and Activation Source Terms in Tokamak Vacuum Chamber Components
During deuterium-tritium (D-T) fusion reactions, the reactor core predominantly generates high-energy fusion neutrons. These neutrons diffuse outward, exposing in-vessel components of the Tokamak vacuum chamber—such as the first wall, blanket, and divertor—to an intense neutron field characterized by high flux and energy deposition. Interactions between neutrons and structural materials within the vacuum chamber induce secondary activation products, which constitute the primary radiation source during reactor shutdown maintenance and serve as critical inputs for radioactive waste classification.
This study employs a Monte Carlo method to investigate the attenuation of neutrons across varying energy levels within the vacuum chamber components, based on a representative Gaussian fusion neutron spectrum derived from D-T reactions. The evolution of the neutron energy spectrum at different radial positions is analyzed, revealing distinct spectral characteristics post-attenuation and corresponding variations in activation source terms across radial distances. Notably, the relationship between activation-induced radioactive source terms and neutron fluence levels is not strictly linear. In fusion environments, where neutron flux and energy are significantly higher, activation source terms at varying depths cannot be derived through simple linear scaling factors.
By selecting representative material compositions of vacuum chamber components, this work utilizes an activation analysis program based on generalized TTA (Transmutation Trajectory analysis) method and the EAF (European Activation File) nuclear database to quantify activation sources under different neutron fluence conditions. Furthermore, the decay dynamics of activated nuclides post-shutdown are systematically investigated.
The results provide critical insights for optimizing material selection for vacuum chamber components, designing radiation shielding strategies during equipment handling, and establishing robust protocols for radioactive waste source-term analysis and classification in fusion reactors.Speaker: yaxiao wang (China Nuclear Power Engineering Co., Ltd.) -
26
Development and Testing of Hybrid HTS Conductor for TF and CS Coils
In future tokamak reactors higher magnetic field will be required to confine the plasma for a longer time to achieve higher confinement density. It is anticipated that higher magnetic field for future tokamak will be provided by high temperature superconducting coils, with good stability and low cryogenic requirements. Second-generation REBCO (2G HTS) high temperature superconductors (@77 K), such as GdBCO, offer high critical current densities and withstand higher magnetic fields compared to conventional low-temperature superconductors (@4 K). A Hybrid High-Temperature Superconductor (HHTS) was developed by soldering five HTS strips (GdBCO, 0.15 mm thick, 4 mm wide) with no insulation (NI) into a copper stabilizer. Five NI-HTS strips without stabilizers were also tested at room temperature and in liquid nitrogen ($LN_{2}$). Experimental results showed that the standalone HTS tapes achieved a maximum stable current of 270 A in LN2 at a voltage of 0.08 V, corresponding to a critical electric field (Ec) of 14μV/mm. In contrast, the HHTS configuration carried a significantly higher current of 740 A at only 0.2 V, with an electric field of 6.8 μV/mm. Numerical simulations using the H-formulation were conducted across the HHTS cross-section, exploring current levels from 500 A to 1000 A and stabilizer widths of 7 mm, 9 mm, and 12 mm. The results indicate stable current conduction up to 1000 A. The copper matrix provides stability during interruption in cooling system, and the electric field in the HHTS allows operation at higher currents. The copper matrix within the conductor enhances stability during cooling interruptions, while the electric field in the hybrid conductor enables operation at higher currents. This configuration provides stable operation at 500 A and above, offering advantages in voltage control and cooling efficiency. Such conductors can be used for the Toroidal Field (TF) coils of Tokamak where long pulse stable current is required.
A small central solenoid (SC) has been manufactured by winding and soldering the 10 layers of HTS strips at copper cylinder (used as a stabilizer). The testing of the pulse current will be performed and results will be reported.Speaker: Zahoor Ahmad (Pakistan Tokamak Plasma Research Institute) -
27
Hydrogen Isotope Behavior in Tungsten and RAFM Steels
The transport and retention of hydrogen isotopes is of vital importance for the realization of future commercial fusion reactors, because it is closely related to plasma operation, fuel recycling and radiation safety. However, the hydrogen isotope behavior in fusion reactor materials is not well understood. Reduced activation ferritic-martensitic (RAFM) steel and tungsten (W) are promising candidate structural materials and first wall materials, respectively. In this presentation, we reported the deuterium permeation and retention behaviors of 7 kind of RAFM steels and 2 kind of W. Gas-driven permeation (GDP) method was used to investigate the deuterium permeability, diffusivity, and solubility of the studied materials, with loading pressure up to 1×105 Pa. For RAFM steels, the results indicates that the deuterium permeability has little materials dependence. In contrast, the deuterium diffusivity of 7 studied RAFMs showed significant variation. And W show lower permeability and diffusivity than RAFMs in the working temperature range. The influence of low-energy high-flux plasma irradiation on deuterium permeation in W has also been studied. Besides, thermal desorption spectroscopy (TDS) was performed to assess the retention behaviors of RAFMs and W with a temperature ramping rate of 0.5 K/s following a static thermal gas charging. Dominant desorption peak of ~1000 K was observed for W while ~500 K for RAFMs. And the retention of deuterium in RAFMs is about 1 to 2 order of magnitude higher than W. Finally, microstructural features contributing to the desorption and retention properties was discussed.
Speaker: Prof. Minyou Ye (University of Science and Technology of China) -
28
Overview and updates of the Burning Plasma Experimental Superconducting Tokamak (BEST)
The Burning Plasma Experimental Superconducting Tokamak (BEST) is under construction in Hefei, China. Designed as a compact high-field tokamak, BEST aims to explore burning plasma physics, achieve advanced steady-state plasma performance (Q>1) , and develop high-Q fusion operations and realization of real-time T production, extraction and cycling. With long pulse duration and compact size, BEST will have high neutron fluence, high density and flux of energy and particles, which provide a test bed for some challenge issues of a reactor: i) Material and PWI physics; ii) Test blanket module; iii) Power and particle exhaust of D&T burning plasma. The physic design has been completed three years ago, and the main parameters has been fixed with Major radius: R=3.6m, Minor radius: a=1.1m, Toroidal Bt=6.15T, Plasma Current:IP=4~7MA, Power: Pfusion=20~200MW, Elongation: κ = 1.7- 1.9, fusion power of 20-200 MW and Q=1-5. The construction of the BEST campus started on 30th June 2023, the main building will be completed in 31st June 2025 and start the pre-assembly of sectors of BEST. The construction of the BEST device is progressing steadily as planned and is going to be completed by the end 2027. With lower cost (both economic cost and time cost), BEST will accelerate the fusion research toward burning plasma. Together with ITER, BEST will strengthen the physics basis and reduce scientific risk of the CFEDR and other future reactors.
Speaker: Prof. Defeng Kong (Institute of Energy, Hefei Comprehensive National Science Center (Anhui Energy Laboratory)) -
29
Quasi-3D thermal–hydraulic analysis and test result of the cool-down for CFETR CSMC
The cool-down of large-scale Cable-in-Conduit Conductor (CICC) superconducting magnets presents significant complexity, requiring careful management to ensure operational safety. This study proposes a quasi-three-dimensional thermal-hydraulic analysis model, specifically designed for large-scale CICC superconducting magnets. The model is applied to the cool-down (300 K-80 K) for the Central Solenoid Model Coil (CSMC) of China Fusion Engineering Test Reactor (CFETR). It is capable of successfully constraining the maximum temperature differential among the various components of the magnet while providing an accurate depiction of the temperature distribution throughout the cool-down. The analysis results correlate remarkably well with the experimental findings. The results offer valuable theoretical insights for future cooling experiments, thereby helping to mitigate the risk of damage to the magnet due to inappropriate cooling parameters.
Speaker: Xiaohui Guan (Chinese Academy of Sciences Hefei Institutes of Physical Science Institute of Plasma Physics) -
30
recent progress on neutronics and tritium analysis for fusion reactor blanket systems in SWIP
.
Speaker: Shen Qu
-
22
-
Tritium Fuel Cycle
-
31
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)
-
31
-
15:20
Coffee Break
-
Tritium Fuel Cycle
-
32
Discussion on key points and session summarySpeakers: Christian Day (Karlsruhe Institute of Technology), Rachel Lawless (UKAEA)
-
33
SummarySpeaker: Christian Day (Karlsruhe Institute of Technology)
-
32
-
16:30
Coffee Break
-
Special Topic
-
34
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) - 35
-
36
Korea’s Strategy for Fusion Energy Realization
Korea’s Strategy for Fusion Energy Realization
Keeman Kim1 and Yeongkook OH2
1Korea Institute of Energy Technology, Naju, 58330, Republic of Korea
2Korea Institute of Fusion Energy, Daejeon, 34133, Republic of KoreaKorean 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. As a following step for the DEMO R&D, the 20th National Fusion Energy Committee has approved the “Strategy for Accelerating Fusion Energy Realization” in 2024 and a new plan is under development. The new plan is based on a Public-Private-Partnership(PPP) and is focused on key technology developments and a design and construction of Compact Pilot Device (CPD). The objectives of the CPD are 1) Construction of innovative compact fusion pilot device in a decade, 2) Production of steady-state capable recirculating fusion power in a decade, and 3) Demonstration of ignited plasma with internal heating only in a decade. The current status of the program will be presented.
Speaker: Keeman Kim (Korea Institute of Energy Technology) - 37
-
34
-
19:00
Banquet
-
-
-
Neutronics
-
38
Chair IntrodcutionSpeaker: Saerom Kwon (National Institutes for Quantum Science and Technology)
- 39
-
40
JA fusion neutronics activities toward JA-DEMO design and construction
Kentaro Ochiai, Saerom Kwon, Yoji Someya, Yoshiteru Sakamoto, Takanori Hirose, Yoshinori Kawamura, Yuki Koga, Kai Masuda, Keitaro Kondo, Kouki Kumagai
National Institutes for Quantum Science and Technology
ochiai.kentaro@qst.go.jpToward the development of JA-DEMO reactors and their neutron engineering are being conducted in Japan. In this talk, we will review the nuclear analysis of the ITER TBM, the current status of the nuclear analysis and design activities of neutron sources for material irradiation, and the progress of nuclear data evaluation, which is essential for fusion, as research activities related to fusion neutron engineering being conducted in Japan.
In JA-DEMO blanket design in Japan, the Quantum Science and Technology Agency (QST) and other organizations are studying the water-cooled ceramic breeding blanket (WCCB) type design as a major candidate, and are conducting analyses mainly on the evaluation of tritium production rate with the ITER reactor configuration in mind.
Irradiation evaluation of structural materials is an important technical study item in DEMO reactor design, and irradiation tests using a strong neutron source such as IFMIF are important for this purpose. The irradiation test analysis of these materials by the D-Li neutron source has been in progress.
Moreover, we performed a few benchmark tests not only for neutron induced nuclear data libraries, TENDL, FENDL and JENDL, but also charged particles induced ones of important nuclides such as iron, copper, concrete and beryllium to improve accuracies for fusion relevant neutronics analyses.
Speaker: Kentaro Ochiai (National Institutes for Quantum and Radiological Science and Technology)
-
38
-
10:35
Coffee Break
-
Neutronics
-
41
Neutronics activities for European DEMO fusion reactor shielding and breeding blanket designs
Dieter Leichtle1, Christian Bachmann2, Jin Hun Park1, Pavel Pereslavtsev2
1Karlsruhe Institute of Technology (KIT), Karlsruhe, Germany
2EUROfusion Consortium, Fusion Technology Department, Garching, Germany
Email corresponding author: dieter.leichtle@kit.eduRecent efforts in the EUROfusion programme towards the realization of a fusion power plant are aiming at developing the conceptual designs of a DEMOnstration fusion reactor and a Volumetric Neutron Source (VNS) facility. The critical role of neutronics and activation studies in support of the design and safety of DEMO requires a strategical approach starting from methodological and validation efforts through development of 3D models to various design and safety related activities. Optimization efforts on conceptual shielding, stipulated by VNS neutronics studies, are supporting the radiological protection framework for DEMO. A comprehensive effort addressing the DEMO breeding blanket designs for the Helium-Cooled Pebble Bed (HCPB), Water-Cooled Lithium Lead (WCLL) and Water-cooled Liquid lead Ceramic Breeder (WLCB) variants entails optimization studies for the respective tritium breeding ratio.
This work has been carried out within the framework of the EUROfusion Consortium, funded by the European Union via the Euratom Research and Training Programme (Grant Agreement No 101052200 — EUROfusion). Views and opinions expressed are however those of the author(s) only and do not necessarily reflect those of the European Union or the European Commission. Neither the European Union nor the European Commission can be held responsible for them.Speaker: Dieter Leichtle (KIT) -
42
Radiation Environment Challenges for a Spherical Tokamak for Energy Production
The Spherical Tokamak for Energy Production (STEP) programme aims to deliver a UK prototype fusion energy plant by 2040, paving the way for commercially viable fusion power. STEP will harness energy from the fusion of Deuterium and Tritium—a reaction that releases 17.6 MeV, with 14.1 MeV imparted to the resulting neutron. To achieve its mission, STEP must meet two critical requirements: (1) demonstrate a net electrical power output of at least 100 MW, overcoming parasitic power loads, which implies a required fusion power of approximately 1.7 GW (∼6x10²⁰ neutrons per second); and (2) attain self-sufficiency in tritium fuel production and processing.
The compact radial design of a spherical tokamak, coupled with the intense neutron environment, introduces several engineering challenges. This work outlines key aspects of the radiation environment and the corresponding design strategies adopted to ensure the plant meets its performance targets. Of particular concern is the protection of radiation-sensitive components, including the High-Temperature Superconducting (HTS) toroidal and poloidal field magnets. The limited inboard space typical of spherical tokamaks renders traditional shielding materials, such as those used in ITER, unable to sufficiently attenuate 14.1 MeV neutrons. As a result, the STEP programme is investigating novel, high-attenuation shielding materials to ensure acceptable magnet lifetimes.
Achieving tritium self-sufficiency requires a Tritium Breeding Ratio (TBR) greater than 1, to compensate for decay, system losses, and tritium retention in materials. Given the spatial constraints of the inboard region, full inboard breeding is unlikely to be viable without compromising magnet lifetimes. Consequently, the burden shifts to the outboard blanket, which must deliver high breeding efficiency, introducing further design complexities.
Additionally, high neutron and gamma fluxes, from both the plasma and activated components such as coolant and structural materials, pose radiation hazards to workers and the public. Mitigating these risks demands accurate dose rate assessments, carefully considered plant layouts, and a maintenance philosophy based around remote handling. These considerations necessitate the development of detailed radiation transport models, pushing the capabilities of current simulation tools. This presentation outlines the radiation workflows developed to model, evaluate, and mitigate these challenges, covering key major radiation sources and their interactions within the STEP plant configuration.
Acknowledgements. This work has been funded by STEP, a major technology and infrastructure programme led by UK Industrial Fusion Solutions Ltd (UKIFS), which aims to deliver the UK’s prototype fusion powerplant and a path to the commercial viability of fusion.Speaker: Tim Eade (UKIFS)
-
41
-
12:25
Lunch Break
-
Posters
-
Neutronics
-
43
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) -
44
OpenMC: a modern, high-performance nuclear design and analysis toolkit for fusion power plants
Aggressive timelines for commercial fusion deployment by the private sector necessitate rapid design iterations to facilitate the build-measure-learn cycle. Many challenges to this process arise in the design of fusion power plants due to the highly integrated nuclear nature of these systems. In particular, nuclear responses throughout the facility are responsible for establishing requirements across many different subsystems; from the safety and shielding requirements of surrounding buildings to the detailed nuclear heating and activation of materials like superconducting magnets, these analyses span the entire lifetime of the facility from design & procurement to construction to operation and decommissioning. This talk discusses the complexities of fusion nuclear analyses and presents the rapidly expanding fusion design capabilities of the open source, community-developed, Monte Carlo code, OpenMC. Over the past five years, many critical features have been added to the code including fixed source transmutation methods, Rigorous 2-Step and Direct-1-Step shutdown dose rate workflows, advanced variance reduction techniques, novel hybrid stochastic-deterministic methods for adjoint transport solutions, and much more. Furthermore, theses tools have undergone, numerous validation studies across many applications, and have been used within uncertainty quantification workflows for determining engineering margins. This talk presents the wealth of recent community-developed features in OpenMC for fusion power plant design and the associated widespread impact across academia, industry, and international organizations.
Speaker: Ethan Peterson (Massachusetts Institute of Technology)
-
43
-
15:15
Coffee Break
-
Neutronics
-
45
Discussion on key pointsSpeakers: Lee Packer (UKAEA), Saerom Kwon (National Institutes for Quantum Science and Technology)
-
46
Summary
-
45
-
Special Topic
-
47
Fusion and Non-ProliferationSpeakers: Lee Packer (UKAEA), Alex Somers (SRNL, USA)
-
47
-
-
-
Closed Session (TPC Meeting)
-
12:30
Lunch Break
-
Tour of Rokkasho Institute
-
