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| Fig. 1: Tokamak Vacuum Vessel. (Source: Wikimedia Commons) |
Nuclear fusion reactor designs have improved over the past many decades, bringing the world closer to commercial viability in the form of Fusion Power Plants (FPPs). The International Thermonuclear Experimental Reactor (ITER), a Tokamak design using deuterium-tritium (D-T) fusion, shows promise in achieving consistent gain greater than 1 (Q > 1). Q > 1 means it will produce more energy than the heating energy required to initiate and maintain the reaction. However, FPPs will not immediately follow the consistent demonstration of Q > 1 by ITER (or even Q = 10, as is anticipated). An intermediate stage of Demonstration Reactors (DEMOs) will be required to test the integrated operation of technologies necessary for net power production; they will ensure a future FPP's sufficient "availability," or its percentage of time generating heat/electricity. [1]
ITER may only reach one full year of plasma operation over its lifetime. DEMOs may only reach 5-10 years of full power operation. Additionally, there is no large-scale fusion representative environment in which to test components at full scale. ITER and DEMOs must therefore refine many operational aspects over their relatively limited lifetimes to ensure the requisite 30-50 years of FPP commercial availability. [2] The known main problems of material damage due to thermal flux and particle activation, as well as remote handling and tritium self-sufficiency, must be iteratively solved by ITER and DEMOs for FPP viability. [1,2]
The sustained operation of an FPP will incur material damage to numerous components due to the extreme heat and particle fluxes of the fusion reaction plasma. In Tokamaks, the vacuum vessel (VV), shown in Fig. 1, houses the plasma. The blanket typically covers the interior components of the VV, shielding them and the superconducting magnets from the plasma. The extent of the damage to blanket materials is highly dependent on their material composition, location, and overall plant design and will directly influence overall FPP efficiency. [3,4]
As an extreme case, the "first wall" of the blanket directly faces the plasma. Plasma-wall interaction varies for different locations of the first wall. A consistent interaction occurs at the divertor, the point for the removal of heat, excess tritium fuel, and the fusion reaction product, helium. The heat is used in electricity generation. The removal of helium "ash" prevents fuel dilution and preserves fusion efficiency. [3,4]
The divertor sustains a high heat load from a large flux of lower-energy particles. This is in contrast to the majority of the first wall, which sustains a moderate heat load from a small flux of high-energy particles. For divertors in FPPs, particle fluxes will be above 1024 m-2 s-1 with a mean particle energy of 5 eV. Temperatures will peak at 1300°C with a mean thermal load up to 15 MW m-2 s-1. [3]
The resultant material erosion and deposition of particles can change the divertor's material properties and reduce its lifetime. Erosion impurities and dust can dilute plasma and reduce its efficiency. Additionally, tritium retention presents a radiation risk. These effects occur in various quantities beyond just the divertor. They extend across the entire first wall, throughout the blanket, and for components behind the blanket. Therefore, material choice and component/plant design are essential to ease incurred maintenance requirements and to increase FPP efficiency. [3,5]
As mentioned above, FPPs will incur planned and unplanned repair, inspection, and maintenance requirements due to material erosion or failure. Remote Handling (RH) of numerous critical systems will be necessary for these requirements due to neutron irradiation of materials and their subsequent emission of gamma radiation. [6] Making RH more difficult are:
Contamination from dust and gaseous tritium
Some level of magnetic field
High levels of heat
Large, heavy, and geometrically complex components with close tolerances (e.g. blanket segments)
Rewelding oddly shaped, neutron-damaged steel
Poor visibility
Distance between the reactor and hot cell [6]
ITER will be equipped with a host of RH equipment. Based off ITER's learnings, DEMOs will augment RH design and procedures to ensure FPPs' sufficient availability. [6]
Self-sufficiency refers to an FPP's ability to breed, process, and use its own tritium while producing sufficient excess for both desired reserves and startup inventory for subsequent FPPs. While deuterium is abundant in the ocean, there are limited natural and man-made tritium resources. [4] Tritium is made as a minor fission product in light water reactors (1 atom generated in 104 acts of fission) or in CANadian Deuterium Uranium (CANDU) reactors. [4,5] CANDU production quantity is expected to peak at ~27 kg in 2027. ITER expects to use 12.3 kg, leaving ~15 kg plus whatever is generated by light water reactors for DEMOs and FPPs. These limitations make FPP tritium self-sufficiency important. [4]
ITER and DEMOs will test various breeding blanket (BB) designs and compositions, where the blanket will both shield reactor components from neutron radiation and use that radiation to breed tritium. [4] To achieve self-sufficiency, BB designs must account for expected breeding reaction efficiency, tritium loss through radioactive decay, and tritium retention and permeation in reactor components. [5]
In numerous designs, lithium in the BB is transmuted into tritium and used as fuel. Beryllium in the BB acts as a neutron multiplier, increasing tritium breeding efficiency by producing two neutrons for every beryllium-neutron reaction. [5] These designs must be validated through actual operation before proceeding with FPP design and construction. [4,5]
Achieving consistent Q > 1 will be a major accomplishment by ITER. However, numerous other challenges must then be overcome to demonstrate future FPP viability. ITER and subsequent DEMOs must solve the problems of material damage, RH, and tritium self-sufficiency prior to the design and construction of FPPs.
© Pete Slye. The author warrants that the work is the author's own and that Stanford University provided no input other than typesetting and referencing guidelines. The author grants permission to copy, distribute and display this work in unaltered form, with attribution to the author, for noncommercial purposes only. All other rights, including commercial rights, are reserved to the author.
[1] C. L. Smith and S. Cowley, "The Path to Fusion Power," Phil. Trans. R. Soc. A 368, 1914 (2010).
[2] R. Kembleton, "Technological Features of a Commercial Fusion Power Plant, and the Gap from DEMO," Fusion Eng. Des. 190, 113544 (2023).
[3] G. A. Cottrell, "A Survey of Plasma Facing Materials for Fusion Power Plants," Mater. Sci. Technol. 22, 869 (2006).
[4] S. Meschini et al., "Modeling and Analysis of the Tritium Fuel Cycle for ARC- and STEP-class D-T Fusion Power Plants," Nucl. Fusion 63, 126005 (2023).
[5] M. Rubel, "Fusion Neutrons: Tritium Breeding and Impact on Wall Materials and Components of Diagnostic Systems," J. Fusion Energy 38, 315 (2018).
[6] R. Buckingham and A. Loving, "Remote Handling Challenges in Fusion Research and Beyond," Nat. Phys. 12, 391 (2016).
