Power and particle exhaust in the ST-E1 fusion power plant
4 affiliations
- 1Tokamak Energy Ltd
- 2Advanced Computing, Photonics & Electromagnetics (CPE), Fondazione LINKS
- 3Oak Ridge National Laboratory
- 4Princeton Plasma Physics Laboratory
Abstract
Power exhaust challenges and potential solutions for a 5 m major radius, low-aspect ratio burning tokamak have been explored. 1D edge plasma models have been used to screen for access to detachment using short and long outer divertor legs in double and single null configurations, using Ar as the primary impurity and assuming tungsten plasma-facing components (PFCs). These show that detachment access can be accessed for all but the most conservative assumptions on scrape-off layer (SOL) width and power, but that trade-offs will be required between magnet engineering and the size of the acceptable window of as-yet uncertain plasma parameters. SOLPS-ITER was used to further model selected plasma scenarios, confirming that Ar seeding can be used to achieve dissipative divertor scenarios with peak deposited heat fluxes below 15 MWm − 2 . Initial scoping of first wall loads and positioning of limiters has been carried out, showing the feasibility of protecting the breeding blanket wall during steady state without impeding tritium breeding. Initial PFC technology selection is also presented, identifying this as a critical area where further work is needed to find an attractive solution for helium-cooled PFCs that can handle high heat fluxes without excessive power requirements. Key questions and trade-offs for concept development have been identified, including: how to achieve high radiation for reduction of SOL power without core performance degradation; whether power exhaust can be well-controlled in a double null plasma; mechanical design and materials challenges of high-heat flux PFCs; and control of material erosion, redeposition and tritium retention.
Related
- Power and particle exhaust for the ARC fusion power plant
- Thermal management and net-power evaluation of the ST-E1 fusion power plant
- Maintenance strategy, structural design, and site layout of the ST-E1 fusion power plant
- Overview of the physics basis for the ARC fusion power plant
- Integrated physics and magnet design for the ST-E1 fusion power plant
BibTeX
@article{robinson2026power,
title = {{Power and particle exhaust in the ST-E1 fusion power plant}},
author = {Matthew Robinson and A. Scarabosio and E. O. Vekshina and Jacob H. Nichols and Jeremy D. Lore and Katarzyna Borowiec and J. Varje and Samara M. Levine and Mattia Scarpari and E.N.J. Maartensson and X. Zhang and Robin Kamenický and J. Alexander and M.D. Jackson and Chris Marsden and Steven McNamara and Rajesh Maingi},
journal = {Nuclear Fusion},
year = {2026},
month = jul,
doi = {10.1088/1741-4326/ae70a0},
url = {https://doi.org/10.1088/1741-4326/ae70a0},
note = {CC BY},
}
RIS
TY - JOUR
TI - Power and particle exhaust in the ST-E1 fusion power plant
AU - Matthew Robinson
AU - A. Scarabosio
AU - E. O. Vekshina
AU - Jacob H. Nichols
AU - Jeremy D. Lore
AU - Katarzyna Borowiec
AU - J. Varje
AU - Samara M. Levine
AU - Mattia Scarpari
AU - E.N.J. Maartensson
AU - X. Zhang
AU - Robin Kamenický
AU - J. Alexander
AU - M.D. Jackson
AU - Chris Marsden
AU - Steven McNamara
AU - Rajesh Maingi
PY - 2026
DA - 2026/07/10
JO - Nuclear Fusion
DO - 10.1088/1741-4326/ae70a0
UR - https://doi.org/10.1088/1741-4326/ae70a0
AB - Power exhaust challenges and potential solutions for a 5 m major radius, low-aspect ratio burning tokamak have been explored. 1D edge plasma models have been used to screen for access to detachment using short and long outer divertor legs in double and single null configurations, using Ar as the primary impurity and assuming tungsten plasma-facing components (PFCs). These show that detachment access can be accessed for all but the most conservative assumptions on scrape-off layer (SOL) width and power, but that trade-offs will be required between magnet engineering and the size of the acceptable window of as-yet uncertain plasma parameters. SOLPS-ITER was used to further model selected plasma scenarios, confirming that Ar seeding can be used to achieve dissipative divertor scenarios with peak deposited heat fluxes below 15 MWm − 2 . Initial scoping of first wall loads and positioning of limiters has been carried out, showing the feasibility of protecting the breeding blanket wall during steady state without impeding tritium breeding. Initial PFC technology selection is also presented, identifying this as a critical area where further work is needed to find an attractive solution for helium-cooled PFCs that can handle high heat fluxes without excessive power requirements. Key questions and trade-offs for concept development have been identified, including: how to achieve high radiation for reduction of SOL power without core performance degradation; whether power exhaust can be well-controlled in a double null plasma; mechanical design and materials challenges of high-heat flux PFCs; and control of material erosion, redeposition and tritium retention.
SN - 0029-5515
SN - 1741-4326
ER -