(38a) Overview of the Pebble-Bed Fluoride-Salt-Cooled, High-Temperature Nuclear Reactor (PB-FHR) and Management of Tritium Air Emissions | AIChE

(38a) Overview of the Pebble-Bed Fluoride-Salt-Cooled, High-Temperature Nuclear Reactor (PB-FHR) and Management of Tritium Air Emissions


Scarlat, R. O. - Presenter, University of California, Berkeley
Huddar, L., University of California, Berkeley
Peterson, P. F., University of California, Berkeley

This presentation provides an overview of the pebble-bed fluoride-salt-cooled, high-temperature class of nuclear reactors (PB-FHRs), and a summary of the supporting experimental and modeling work ongoing at UC Berkeley. This presentation also provides an overview of the challenges associated with management of air emissions of tritium from PB-FHRs. Tritium permeation barriers for the metallic heat exchangers, in addition to tritium extraction system from the fluoride salt, need to be designed.

The PB-FHR concept is currently under development at UC Berkeley , in collaboration with other universities and national laboratories. PB-FHRs are cooled by liquid fluoride salt eutectic mixtures, with a randomly packed bed of heat generating fuel pebbles, which rely on natural circulation for emergency decay heat removal to ambient air. Water and heat transfer oils can be used as liquid salt simulant fluids and they provide the advantage of low distortion thermal-hydraulic experiments at significantly lower temperatures, reduced geometric scale, and reduced thermal power input. The goal of FHRs is to provide an energy technology with a short commercialization timeline and significant safety advantages compared to advanced light water nuclear reactors currently under construction, at costs that are competitive with natural gas power plants.

Tritium is produced in the PB-FHR core from neutron activation of 6Li in the fluoride salt coolant, and all of the isotopes of hydrogen readily diffuse through metallic structures at the reactor operational temperatures of 600 to 700oC. In the PB-FHR, metallic structures in contact with the tritium-carrying salt include transport piping exterior to the reactor vessel, and heat exchangers. The baseline UC Berkeley PB-FHR design uses a nuclear air combined cycle (NACC) for power conversion, and the primary salt is used to directly heat compressed air in coiled tube air heaters (CTAHs). These CTAH heat exchangers have much larger surface area than the transport piping, so they are the dominant leakage path of tritium from the reactor system.

A PB-FHR core produces on the order of 1000 times the tritium emitted as liquid and gaseous effluents by pressurized water reactors (PWRs) operating today in the United States. It is the intent of the PB-FHR designers to maintain PB-FHR normal operation tritium emission rates close to those from PWRs. Thus, an additional sink for tritium must be designed into the PB-FHR salt coolant system, in order to keep the tritium flow rates through the CTAH heat exchangers under one thousandth of the tritium production rate in the core.  Furthermore, based on the current efficiency of tritium-permeation barriers for metallic surfaces, the tritium concentration in the salt must be maintained on the order of parts per billion.



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