Conceptual design for a blanket tritium extraction test stand
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Atmospheric Radiation Measurement (ARM) Data Center
Tritium breeding is fundamentally required for a sustainable fusion fuel cycle, yet the technological readiness of blanket technology lags far behind other fusion systems. Breeder concepts are divided into solid and liquid media, where solid breeders typically rely on a sweep gas, such as helium, to carry away tritium from lithium containing ceramic materials, and liquid breeders produce tritium from lithium containing eutectics (e.g., PbLi) or molten salts (FLiBe). In each case, tritium must be harvested from the breeding medium. A promising method for tritium extraction is through a vacuum permeator, in which a concentration gradient from the tritium- containing fluid promotes diffusion through a membrane with high hydrogen permeability to the vacuum. This technology has been demonstrated for hydrogen gas systems using Pd and PdAg permeators, but relatively little work has been done to test tritium extraction from PbLi. A Tritium Extraction eXperimental (TEX) loop is being designed to test tritium extraction in a vacuum permeator configuration. The system design is such that it will allow the testing of tritium extraction from both helium and PbLi. A phased approach is being taken that will allow testing of small specimens for fundamental permeation measurements, to multi-meter component testing at near-prototypic conditions. A molten PbLi loop is challenging due to the toxic and explosive nature of Pb and Li, respectively, radiological concerns by introducing tritium, and high temperatures involved in such a system. In addition, PbLi corrosion is a significant issue at high temperatures (>400C). The TEX system will not employ a neutron source for volumetric production of tritium. Herein we present the design and methods used for 1) pumping PbLi, 2) introducing deuterium or tritium into the PbLi, 3) quantifying the amount of deuterium in the loop, 4) extracting deuterium and tritium, and 5) quantifying the total amount of extracted deuterium or tritium from the permeator. In addition, the safety design for operating such a system will be discussed.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- Contributing Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- DOE Contract Number:
- DE-AC07-05ID14517
- OSTI ID:
- 1903777
- Report Number(s):
- INL/CON-20-60486
- Resource Relation:
- Conference: Technology of Fusion Engineering (TOFE), November 16-19, 2020, Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83402
- Country of Publication:
- United States
- Language:
- English
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