Tritium aspects of the fusion nuclear science facility
The Fusion Nuclear Science Facility (FNSF) is intended to bridge the technical gaps from ITER to DEMO, and provide a fully nuclear environment for the advancement of materials and blanket science, among other missions. Though smaller (~500 MW fusion power) than the envisioned DEMO, our FNSF design still seeks to test DEMO-relevant components and technologies, and to that end one of its goals is to achieve a tritium breeding ratio near one using a dual-cooled lead lithium (DCLL) blanket. We explore here the migration of bred tritium through in-vessel components and coolant loops, which is exacerbated by the relatively low solubility of tritium in PbLi and the high operating temperatures envisioned in later phases of operation. Some features inherent to the design of the DCLL help limit tritium losses, namely the relatively high PbLi flow rates; coupled with efficient tritium extraction, the DCLL potentially achieves low tritium permeation losses without the use of permeation barriers. We outline here the design of such an extraction system, and some of the materials challenges that must be met in order to realize it. The efficacy of these and other design features, such as coaxial piping, is demonstrated via a series of parameter studies. We also investigate the effect of parameter uncertainties, and find that the imprecise knowledge of the tritium solubility in PbLi in particular leads to very disparate estimates of tritium losses. Depending on the experimental value used, losses range from ~0.05 to over 6 grams of tritium per year. As a result, the former is acceptable based on estimated environmental releases and resultant dose to members of the public; the latter exceeds regulatory limits.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE Office of Science (SC), Fusion Energy Sciences (FES)
- Grant/Contract Number:
- AC07-05ID14517
- OSTI ID:
- 1480304
- Report Number(s):
- INL/JOU-17-41223-Rev000
- Journal Information:
- Fusion Engineering and Design, Vol. 135, Issue PB; ISSN 0920-3796
- Publisher:
- ElsevierCopyright Statement
- Country of Publication:
- United States
- Language:
- English
Web of Science
The Impacts of Liquid Metal Plasma-Facing Components on Fusion Reactor Safety and Tritium Management
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journal | November 2019 |
Critical Exploration of Liquid Metal Plasma-Facing Components in a Fusion Nuclear Science Facility
|
journal | June 2019 |
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