Fauske & Associates, LLC (FAI), Argonne National Laboratory (ANL), and Westinghouse Electric Company are collaborating within the program “Development of an Integrated Mechanistic Source Term Assessment Capability for Lead- and Sodium-Cooled Fast Reactors.” This program, partially funded by the U.S. Department of Energy through the Gateway for Accelerated Innovation in Nuclear initiative, aims at developing a computational framework for predicting radionuclide release from a broad spectrum of accidents that can be postulated to occur at liquid metal cooled reactor (LMR) facilities. Specifically, the program couples the transient and severe accident analysis capability of the SAS4A/SASSYS-1 code developed by ANL with the radionuclide transport analysis capability of the Facility Flow, Aerosol, Thermal, and Explosion (FATE) code developed by FAI. The testing of both the individual codes and of the coupled system is performed on a generic lead cooled fast reactor (LFR) design that is intended to capture the key differences between the LFR and the sodium fast reactor (SFR), around which the SAS4A/SASSYS-1 code has historically been developed and from which the coupled code inherits some features requiring modification before application to LFR systems. By means of this approach, a computational framework applicable to both LFR and SFR systems will be obtained that will assist LMR developers in performing a realistic, scenario-dependent mechanistic source term (MST) assessment expected not only to strengthen their safety case but also to support easier siting and claims on reduced emergency planning zone requirements. This paper discusses the work being performed to adapt the SAS4A/SASSYS-1 and FATE codes to LFR technology; the code coupling method implemented; and some of the results of the LFR test case, with the latter aimed at demonstrating the progress made toward the development of the MST analysis capability that is ultimately targeted.