A new study by the International Atomic Energy Agency (IAEA), Deuterium–Tritium Toroidal Magnetic Confinement Facilities (IAEA-TECDOC-2116), looks at decommissioning and waste management with respect to fusion facilities. Experimental fusion facilities are integral to advancing fusion research and informing the design of future power-producing plants. As some of these facilities enter decommissioning, early experience is being gained in dismantling activated components and managing radioactive waste, including tritiated materials.
This emerging body of experience provides an important foundation for developing the technical methodologies and waste management strategies that will be required for future facilities, where higher neutron fluences and larger tritium inventories are expected. This new publication is intended for fusion researchers and engineers actively involved in radioactive waste management for fusion, as well as decommissioning experts with practical experience and regulatory authorities.
IAEA says the decommissioning of deuterium–tritium (D-T) magnetic confinement fusion facilities has several unique aspects arising from their complex design and mode of operation, with neutron activated components being of particular importance. This reflects the high emission rate of energetic neutrons, which are emitted with energies of about 14 MeV, characteristic of D-T plasmas.
Another important factor is the use and production of tritium during operation. The high mobility of this isotope and its tendency to diffuse into facility components can result in significant tritium inventories and volumes of tritium contaminated materials that need to be managed.
Many experimental fusion facilities around the world have already been operated and subsequently closed, although not necessarily decommissioned. These facilities typically had a low radiological source term because they mostly operated without tritium, generally using hydrogen or deuterium plasmas instead. However, several experimental fusion facilities currently under development or in operation plan to use D-T plasma, primarily because of its lower ignition temperature. These facilities aim to support the development of next generation facilities, including demonstration plants, pilot plants and eventually fusion power plants.
To ensure effective decommissioning, issues related to the high neutron flux environment and tritium usage, particularly in fusion power plants, will need increased attention. Key considerations include workforce radiation exposure during dismantling, tritium handling, remote operations and the management of radioactive waste. There is currently only limited international experience with the dismantling of fusion facilities that have operated with D-T plasmas, although valuable insights have been gained from the dismantling of experimental facilities such as the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory in the US. Additional experience is expected to be gained in the coming years from the dismantling of the Joint European Torus (JET) in the UK.
The overall dimensions of the international ITER facility, currently under construction in France, as well as those of other fusion facilities being planned, are significantly larger than those of TFTR and JET. The systems, structures and components that make up these large facilities are expected to generate substantial quantities of radioactive waste upon dismantling, owing to both material activation and tritium contaminated components and materials, which will necessitate effective safety measures.
The new TECDOC sets out technical considerations, presents lessons learned and provides IAEA member states with other practical information relevant to the future decommissioning and waste management of toroidal magnetic confinement fusion facilities, especially those involving D-T operation.
It reports on experience gained and current good practice regarding planning and implementation of decommissioning of D-T toroidal magnetic confinement fusion facilities, including aspects relevant to the management of the resulting materials and radioactive waste. Regulatory considerations will become increasingly important as fusion facilities grow in complexity and fusion technology moves closer to deployment.
While some member states are developing regulatory frameworks for the operation and eventual decommissioning of such facilities, global experience with regulating D-T fusion facilities remains limited. Even if there are differences between the various national regulations, decommissioning planning will need to consider the following main phases and associated administrative steps:
Pre-shutdown and post-shutdown phases. An important activity during this period is the submission of the decommissioning plan to the appropriate regulatory authorities.
Depending on the decommissioning strategy, the facility will prepare for: immediate dismantling or deferred dismantling; and termination of the authorisation – conducting a final survey, which is then verified by the regulatory authority. The results of this survey determine whether any restrictions need to be applied to the future use of the site or if it can be released for unrestricted use.
Current global experience in regulating fusion-generated radioactive waste comes primarily from experimental fusion facilities such as TFTR (USA), JET (UK), and others. Experience from the decommissioning of experimental D-T magnetic confinement fusion facilities has shown that such programmes can be lengthy and complex, primarily due to the scale and technical intricacies of the facilities. As a result, funding mechanisms for decommissioning and waste management need to be resilient to uncertainties and associated risks. These mechanisms need to be established prior to the start of operation, ensuring that adequate financial resources are allocated throughout the facility’s operating lifetime.
Decommissioning considerations for fusion facilities need to be integrated from the design stage to ensure that the management of radioactive waste and the eventual decommissioning can be carried out safely and efficiently. Key aspects include:
Designing for decommissioning, including modularisation of plant components, facilitating easy access to systems, structures and components, and minimising the use of materials that may complicate dismantling.
Selecting structural material with low impurities so that the extent of future activation is minimised (both in terms of volume and overall radiological activity).
Design solutions need to aim to manage tritium inventory in radioactive waste and in structures, systems, and components. This include selecting materials compatible with detritiation treatments, designing components to fit within detritiation furnaces, and easy enabling segregation of materials with different detritiation requirements. Key technical lessons from the decommissioning of TFTR and JET include the use of specialised processes such as remote handling, robotics, volume reduction techniques and detritiation ovens to recover and reduce tritium in materials while minimising radiation exposure to personnel.
For managing radioactive waste from fusion facilities, particularly those using D-T plasma, key considerations include:
Pursuing the potential for decontamination, reuse, or clearance of mildly radioactive materials as a means of minimising, to the extent practicable, the volume of waste requiring disposal in radioactive waste repositories.
Undertaking research to assess the feasibility of recycling highly radioactive metal-based components, along with cost-benefit analyses comparing options with direct disposal.
Materials selection and design optimisation: careful selection of materials with low impurity levels can significantly reduce the volume and activity of radioactive waste as well as environmental impacts such as liquid discharges.
Processing radioactive waste to remove or reduce specific radionuclides can help reduce overall waste volume and activity, improving the prospects for acceptance at disposal sites. Research is ongoing into techniques such as annealing and melting, which aim to remove tritium and reduce carbon-based activation products in contaminated or activated materials.
In fusion, much of the external radiological hazard at shutdown can be effectively managed through decay storage or deferred dismantling over a period from five years to many decades. However, this approach carries the risk of knowledge loss over time, which needs to be mitigated through robust documentation and information management strategies.
As designated solutions for managing waste from the operation and decommissioning of fusion facilities, including storage and disposal, need to align with national legislation, it is important that such legislation adequately addresses the specific needs of these facilities. This may include considerations on how the waste acceptance criteria of existing repositories can be met, particularly in cases where the quantity or types of radionuclides differ significantly from those anticipated in the original safety cases for those repository facilities.
Finally, several areas of active research and development have been identified as critical to enhancing future decommissioning efforts and addressing radioactive waste management needs. These include:
Remote handling cutting techniques capable of addressing the complexity of dismantling the tokamak core, with a focus on minimizing the spread of particulate contaminants within the facility and reducing the generation of secondary waste.
Waste packages need to be developed that can efficiently confine tritium and reduce outgassing: specific matrices and leak tight sealings.
Radiological characterisation tools (spectrometry, sampling) need to be developed and adapted to activated large scale components and be capable of determining tritium content with less intensive methods than destructive sampling.
Calculation tools or models need to be developed to estimate tritium diffusion into materials.
Industrialisation of waste detritiation techniques such as baking or melting need to be developed and adapted to handle large volumes and sizes of materials from vacuum vessel components.
Technologies ought to be developed with long-term maintainability in mind and designed to be easily understood by future generations of operators and stakeholders.
