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The Divertor Tokamak Test (DTT) facility will start its operations in 2026. DTT will operate with D-D fuel only, for an expected operational period of 25 years. Nevertheless, tritium will be produced by the D(d,p)T reaction. A mandatory step in the safety assessment of the machine is the estimation of the different source terms. Major contributions to the source terms are due to tritium and to activated dust. The amount of tritium in the vacuum chamber, in co-deposited tungsten layers and implanted in the bulk of the first wall is computed in this work. Also, a preliminary estimation of dust production due to inter and intra ELMs sputtering is carried out. Results report small amount of source terms related to tritium, below 1 mg after one year of full power operations, and less than 300 g of activated dust at the end of life.

The new Affordable Robust Compact (ARC) fusion reactor, which, compared to larger machines like ITER, aims to achieve its goal of fusion energy in a less expensive and smaller but even more...

The International Energy Agency (IEA) has been promoting the IEA Environment, Safety and Economic Aspects of Fusion Power program for many years. Among the tasks of this program, one task in partic...

The fusion pilot power plant ARC is a conceptual design of a D–T Tokamak under investigation at the Massachusetts Institute of Technology. Special attention is paid on the radiological hazard, which until now has been translated in the reduction of materials activation. Indeed, one of ARC main goals is to be fast deployable in any US site: thus, the radiological risk associated to its presence must be minimized, both for the population and the environment. Tritium is one of the main sources of radiological hazard in ARC and it is almost ubiquitous: it is found in the vacuum chamber, in the blanket, in structural materials and in tritium processing and storing components. In this work, a safety analysis is proposed to quantify the radioactivity release following an accidental scenario. Tritium inventories in the main components are estimated starting from the preliminary design of the FLiBe circuit. The source term is quantified assuming the occurrence of a severe accident damaging key components. Afterward, the environmental impact and the doses to the most exposed individuals are evaluated through suitable population doses codes, and ARC compliance with safety limits is assessed.

Fusion waste will not contain transuranics and fission products like fission. Other than volumes and total radioactive inventories, a comparison among radioactive material production in different energy sources should take into account the different nature of radioactive nuclides, in terms of their hazard potential. A convenient comparison can be made with the total radiotoxicity indices. We have performed such a comparison considering fusion power plant models, GEN II PWR and GEN IV fission reactors, and ashes from a coal-fired power plant, than bear naturally occurring radioactive materials. The results are normalized, and total indices have been calculated considering the complete materials and fuel cycle for the reactors. Fusion compares favorably with fission, as expected. If low activation materials are used, fusion radiotoxicity indices, after a cooling time of the same order of magnitude as the interim storage envisaged, are also even lower than that of coal ashes.

CANDOR is a high-field advanced fusion fuel cycle experiment based on Ignitor, but with larger dimensions and higher fusion power: it is a feasibility study of a high-field Deuterium-Helium-3 (D3He) experiment of larger dimensions and higher fusion power than Ignitor, still based on the core Ignitor technologies. Results of investigations on the feasibility of D3He burning and side neutrons production in D3He plasmas and specifically in CANDOR show that, with the initial use of DT triggering, the need for an intense auxiliary heating would be considerably alleviated. The total released 14 MeV neutron energy during the 16-second burning sums to about 210 MJ. DT and DD neutron currents incoming in the CANDOR plasma chamber wall and the Neutron Wall Loads have been computed. D3He ignition could be studied in CANDOR, with modest and conservative developments of the present technology. CANDOR has a low neutron wall loading, softer neutron spectrum, low radiation damage, and - consequently - lower neutr...

At ENEA Brasimone Research Centre, Italy, a new experimental facility named TRIEX-II ( Tri tium Ex traction) was designed and installed. Its aim is to characterize, in the range of operating conditions foreseen for the European Test Blanket System WCLL-TBS, several extraction technologies for hydrogen isotopes (Q2) solubilized in the flowing metallic LiPb alloy (15.7 at. % Li). One of these technologies is the packed column, an example of Gas/Liquid Contactors (GLCs). This paper proposes a multiscale modelling tool, combining different scales through two computational tools. The extraction column mock-up is described by a component-detail level model, developed in COMSOL Multiphysics, and integrated into a system level code of the whole TRIEX-II circuit, developed using MATLAB/Simulink. The integration is carried out by implementing the COMSOL component into an S-function of MATLAB/Simulink, preserving the process flow diagram of the loop. In this way, it was possible to quantify the Q2 concentrations and the permeation fluxes, and to evaluate the theoretical extraction efficiency. The LiPb flow field inside the extractor was also derived. Finally, a comparison with the experimental results was performed. The results suggest that this tool could be adapted for analyses of complex systems, at a multiscale level, in view of design improvements and safety studies for the tritium cycle of ITER.

In the worldwide energy industry, nuclear fusion could be a breakthrough in the medium-long term. One promising fusion machine under design at Massachusetts Institute of Technology is ARC reactor. It is likely that the first nuclear fusion plants will rely on a traditional thermodynamic cycle for the downstream power energy conversion. In this framework, one of the design aspects is to maximize the thermal efficiency. In the present paper the thermodynamic cycles, which could be adopted in ARC rector, are explored. Three cycles have been considered: the Rankine, the Brayton and a combined cycle. For the gas adopted in the Brayton and combined cycles, two options have been investigated: supercritical Helium and supercritical CO2. A comparison among thermal efficiency and preliminary considerations on component integrity’s, plant feasibility and economics of each studied configurations has been discussed to identify the possible best option for ARC reactor. The results show that a regenerative CO2 Brayton cycle with intercooler and re-heating systems is the most promising one. Such configurations is able to reach a thermodynamic efficiency of up to 0.6.

The Water-Cooled Lithium-Lead (WCLL) is one of the four breeding blanket concepts proposed by Europe in view of its DEMO reactor. The velocity field of the electrically conducting lead-lithium eutectic alloy inside the blanket is strongly influenced by the external magnetic field used for plasma confinement combined with buoyancy effect. The strength of the magnetohydrodynamics (MHD) effect and of the magneto-convective effect (MHD and buoyancy force) depends on the intensity of the magnetic field and its orientation with respect to the direction of the lead-lithium motion. This phenomenon significantly influences the resulting temperature and velocity fields, and therefore the tritium transport inside the breeding blanket. A multi-physics approach of a 3D tritium transport model is presented for a simplified geometry of the WCLL breeding blanket. In particular, advection-diffusion of tritium into the lead-lithium eutectic alloy, transfer of tritium from the liquid interface towards the steel, diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant, and advection-diffusion of diatomic tritium into the coolant, temperature field, velocity fields of both lead-lithium and water, buoyancy forces, and MHD effect have been included in this study. The tritium concentrations and the inventories inside the lead-lithium, in the Eurofer pipes and in the baffle, and in the water coolant have been evaluated.

The proposed blanket for Affordable Robust Compact (ARC) reactor is one of the simplest blanket concepts. It is a bulk tank filled with a lithium and beryllium fluorides molten salt. The fluid effectively works as tritium breeder, vessel coolant and neutron moderator and shield. However, despite the simplicity of the concept, the compactness of the reactor constitutes a novelty in the fusion field. It is thus necessary to evaluate all the possible solutions for an effective blanket component. This work analyses different liquid blanket identifying the most suitable for a compact fusion reactor. More specifically, the study addresses the capability of breeding tritium in a compact solution, actively shielding the coils and reducing the radioactive waste. Findings are that FLiBe optimizes the most the system in terms of applicability, tritium breeding, compactness and activation. Nonetheless, there is no lack of backup choices. For instance, there are hints that lithium-zirconium fluoride salts could accomplish the blanket main tasks in a compact reactor too. Leaving PbLi as inefficient, but cheap and still virtually viable solution.

Affordable Robust Compact reactor is a conceptual design for a Tokamak conceived by Massachusetts Institute of Technology (MIT) researchers. The design of this tokamak is under development and update. One of the key parameters for fusion reactor power plants is the tritium breeding ratio (TBR), which has to guarantee the tritium self-sufficiency. The tritium inventory circulating in a fusion power plant must be minimized. In the meantime, to enhance plant’s economics, the amount of tritium generated and stored should be maximized, since it would be used to startup new reactors. Both of the aforementioned trends meet their best in a TBR as high as possible. In this work, ARC tritium breeding ratio is studied and optimized. Taking advantage of Monte Carlo neutron transport codes, several configurations of ARC’s blanket and vacuum vessel have been analyzed in order to find the most effective one for a high TBR. The study takes into account different materials for the structure, such as Inconel718, V-15Cr-5Ti and Eurofer97. Moreover, it scans different width of coolant’s channels and evaluates the effect of lithium-6 enrichment in the blanket looking for the best configuration in terms of TBR.

The Water-Cooled Lithium Lead (WCLL) breeding blanket is one of the European blanket designs proposed for DEMO reactor. Tritium can permeate into the different structural materials, arising potential issues concerning the fuel self-sufficiency and can be lost into the environment with consequent radiological hazard for the population. Within this frame, a tritium transport analysis is fundamental to evaluate tritium retention in LiPb (15.7 at. % Li) and in the structures and tritium permeation fluxes into the cooling water. To assess this study, a portion of the breeder unit of the outboard equatorial module of the WCLL was modelled. The buoyancy forces and the magnetohydrodynamic (MHD) effect were also included. The final system of partial differential equations was solved with a novel approach through COMSOL Multiphysics. The coupled MHD and heat transfer system of equations was solved performing a transient simulation, that was stopped when the main average variables, temperature and velocity, reached a stable condition. In this way, it was possible to determine the lithium-lead velocity field and to use it as an input for the transport analysis. Tritium transport was modelled by using the input data of tritium generation rate and volumetric power deposition coming from an ad-hoc Monte Carlo simulation realized with MCNP software. Moreover, the transport analysis included advection-diffusion of tritium into the LiPb, transfer of tritium from the liquid interface towards the steel, diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant, advection-diffusion of diatomic tritium into the coolant.

The Affordable Robust Compact (ARC) reactor is a conceptual design for a Tokamak conceived by Massachusetts Institute of Technology researchers. The ARC design is under development and update. Since ARC will be a D–T tokamak, neutron generation and material activation will be main issues for safety studies and assessment of environmental impact and siting questions. The safety assessment goal for ARC is to demonstrate that it could be easily sited in the US, without public health and environmental problems and the need of any emergency plan implying population evacuation or sheltering. Another safety feature that will be verified is the need of a containment building in which the reactor should be surrounded. Starting from activation studies already developed for the ARC’s vacuum vessel structure and the liquid blanket as well, a further and deeper analysis, that includes the first wall and neutron multiplier layer activation, has been carried out. Afterwards, taking advantage of the RESRAD population dose code, the study arrives to the assessment of doses to most exposed individuals from accidental activated material release in atmosphere, including possible tritium releases: radioactive safety limits for ARC environmental impact are finally defined.

Neutron irradiation is one of the most concerning issues to design plasma facing components and reactor inner structures in fusion devices, especially for high power density ones, like ARC reactor. This study addresses the main aspects of neutron irradiation on solid materials of ARC reactor. In particular it deeply analizes the effect of neutron induced activation proposing low activation structures, like vanadium alloys and different optimization methods like isotopic tailoring, detritiation and impurity control. Furthermore, irradiation damage issues and their dependence on the energy spectrum are highlighted. It resulted that V-Cr-Ti alloys dramatically reduce the radioactive inventory of ARC with respect to its baseline configuration, which proposes the application of Inconel 718. Such alloy is also optimizable through the tailoring of titanium isotopes and is virtually capable of hitting recycle limits in a couple of decades. Lastly, it shows a relatively growth of gas during irradiation. However, it is highlighted how experiments on neutron damage featuring fission neutrons risk to be able to tell very little about the behavior of the same materials under fusion neutrons, as damaging mechanisms seem to be different.

Nowadays, Fusion Energy is one of the most important sources under study. During the last years, different designs of fusion reactors were considered. At the MIT, an innovative design was created: ARC, the Affordable Robust Compact reactor. It takes advantage of the innovative aspects of recent progress in fusion technology, such as high temperature superconductors, that permit to decrease the dimensions of the machine, reaching at the same time high magnetic fields. Our main goal is the low-activation analysis of possible structural materials for the vacuum vessel, which is designed as a single-piece placed between the first-wall and the tank that contains the breeding blanket. Due to its position, the vacuum vessel is subject to high neutron flux, which can activate it and cause the reduction of the component lifetime and decommissioning problems. The activation analysis was done also for the liquid breeder FLiBe, compared with Lithium-Lead. Codes used for the low-activation analysis were MCNP and FISPACT-II. The first one is based on a neutronics model and for each component a certain neutron flux is evaluated. For FISPACT-II, the main input is the composition of the analyzed material, the neutron flux and the irradiation time. Results from FISPACT-II are the time behavior of specific activity, contact dose rate. To assess suitable structural materials for the vacuum vessel, low-activation properties were considered. Vanadium alloys turn out to be one of the best alternatives to the present material, Inconel-718. Finally, isotopic tailoring and elemental substitution methods were applied. Here, the composition of each alloy is analyzed and critical isotopes or elements are eliminated or reduced. After the modifications, new simulations are done, and those leading to significant improvements in the final results are highlighted.

The Helium-Cooled Lithium Lead (HCLL) breeding blanket is one of the European blanket designs proposed for DEMO reactor. A tritium transport model is fundamental for the correct assessment of both design and safety, in order to guarantee tritium self-sufficiency and to characterize tritium concentrations, inventories and losses. The present 2D transport model takes into account a single breeder unit located in the outboard equatorial module of the HCLL breeding blanket, which is one of the most loaded modules in normal operating conditions. A multi-physics approach has been adopted considering several physics phenomena, providing for buoyancy effect, temperature fields, tritium generation rate and velocity profile of lead-lithium and coolant. The transport has been modelled considering advection-diffusion of tritium into the lead-lithium eutectic alloy, transfer of tritium from the liquid interface towards the steel (adsorption/desorption), diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant (recombination/desorption), advection-diffusion of diatomic tritium into the coolant. Tritium concentrations, inventories and losses have been derived under the above specified phenomena. In particular, the effect of buoyancy forces on the tritium transport has been implemented and compared with the condition without buoyancy.

The Helium-Cooled Lithium Lead (HCLL) breeding blanket is one of the European blanket designs proposed for DEMO reactor. A tritium transport model is fundamental for the correct assessment of both design and safety, in order to guarantee tritium self-sufficiency and to characterize tritium concentrations, inventories and losses. The present 2D transport model takes into account a single breeder unit located in the outboard equatorial module of the HCLL breeding blanket, which is one of the most loaded modules in normal operating conditions. A multi-physics approach has been adopted considering several physics phenomena, providing for buoyancy effect, temperature fields, tritium generation rate and velocity profile of lead-lithium and coolant. The transport has been modelled considering advection-diffusion of tritium into the lead-lithium eutectic alloy, transfer of tritium from the liquid interface towards the steel (adsorption/desorption), diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant (recombination/desorption), advection-diffusion of diatomic tritium into the coolant. Tritium concentrations, inventories and losses have been derived under the above specified phenomena. In particular, the effect of buoyancy forces on the tritium transport has been implemented and compared with the condition without buoyancy.

The Helium-Cooled Lithium-Lead (HCLL) and Water-Cooled Lithium-Lead (WCLL) Breeding Blankets are two of the four European blanket designs proposed for the DEMO reactor. A tritium transport model inside the blanket is fundamental to assess its preliminary design and safety features. Tritium transport and permeation are complex phenomena to be taken into account in the evaluation of the tritium balance, in order to guarantee tritium self-sufficiency and to characterise tritium concentrations, inventories and losses. In this context, the study has been performed at the minimal functional unit level of the outboard equatorial breeding blanket module, which is, during continuous operative conditions, one of the most loaded modules and this results in higher permeation phenomena. For these purposes, a 2D model for the breeder unit of HCLL and a 3D model for the single cell of WCLL breeding blanket concepts have been investigated. The models include advection-diffusion of tritium into the lead-lithium eutectic alloy, transfer of tritium from the liquid interface towards the steel, diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant, and advection-diffusion of diatomic tritium into the coolant. Thermal field, tritium generation rate profile, velocity field of lead-lithium and coolant have been also taken into account.

A safety analysis study has been applied to the Ignitor machine. Deterministic evaluations of radioactive environmental releases have been performed. The preliminary radiological impact analysis for the normal operation and the main accidental sequences of Ignitor, in case of its localization in the TRINITI site in Russia are presented. The site of TRINITI, hosting since decades nuclear installations, is well characterized, both from the meteorological and population aspects: many data have been collected over the years. The Ignitor machine, both during routine functioning and accidental sequences, presents a negligible radiological impact. Ignitor does not need any concrete primary containment, like those used in fission nuclear power plants. There is no need of people evacuation or emergency countermeasures even in presence of the worst accident.