TRITIUM CONTROL and CAPTURE in SALT-COOLED FISSION AND FUSION REACTORS: EXPERIENTS, MODELS, and BENCHMARKING

Report Date: January 2016
Appendices: No

Abstract

The Workshop on Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Experiments, Models and Benchmarking was held on October 27-28, 2015 in Salt Lake City. The workshop objectives were to bring together researchers involved in experiments, modeling and benchmarking for tritium control at ~700°C in liquid salts and related systems to (1) exchange information and enable the future exchange of information, (2) initiate an effort for benchmarking of experiments and models, and (3) encourage cooperation between different groups working on the same challenges.

The workshop was organized by 5 organizations with common interests in tritium in salt systems at high temperatures: the Department of Nuclear Science and Engineering (NSE) at the Massachusetts Institute of Technology (MIT), the Nuclear Reactor Laboratory (NRL) at MIT, the Plasma Fusion Center (PFC) at MIT, the University of Wisconsin at Madison (UW) and the Chinese Academy of Science (CAS).

These diverse organizations have a common interest because three advanced power systems use liquid salt coolants that generate tritium and thus face common challenges. The Fluoride-salt-cooled High-temperature Reactor (FHR) uses the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors and fluoride salt coolants. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium to the salt. In both systems, the base-line salts contain 7Li. Isotopically separated lithium is used to minimize tritium production. The Chinese Academy of Science plans to start operation of a 10-MWt FHR and a 2-MWt MSR by 2020. High-magnetic field fusion machines proposed to use lithium enriched in 6Li to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of lithium salts as coolants.

This proceedings summarize results from that workshop including descriptions of the power systems that use high-temperature salts, the common chemistry and tritium challenges, ongoing work removing tritium using carbon, other technologies for tritium control, and other tritium capabilities in the U.S. The appendixes include the workshop agenda, participants, and presentations from the workshop.

Program:     ANP : Advanced Nuclear Power Program

Type:     TR 

RPT. No.:  165