Nuclear Science and Engineering

Simplified Radionuclide Transport Source Term Analysis Code

 SRT
Tool for Sodium Reactor and Microreactor Radionuclide Transport and Retention Analysis

SRT is a computational tool developed by Argonne National Laboratory for rapid mechanistic source term analyses for pool-type, metal fuel SFR designs and metal fuel microreactor designs. SRT provides time-dependent radionuclide inventories throughout the reactor system and building, along with offsite dispersion and dose results using Gaussian plume models. The code is designed to facilitate uncertainty and sensitivity analyses for the determination of factors and phenomena influential in the transport and retention of radionuclides.

Features

  • Highly-customizable models through user input
  • Uncertainty/sensitivity parameters within user input
  • Models for multiple radionuclide transport/retention phenomena:
    • In-pin radionuclide migration
    • Release from failed fuel
    • Bubble transport through the sodium pool
    • Vaporization from the sodium pool
    • Radionuclide behavior in the cover gas region or inert reactor vessel
    • Radionuclide behavior in the reactor building compartments
    • Offsite dispersion calculation
  • Customizable reactor building compartments
  • Radionuclide release from ex-core sources
  • Detailed graphical and text outputs

How to obtain the code

SRT is available for licensing. Please contact SRThelp@​anl.​gov for further information.

Recent Publications

Outline of SRT​’s sodium fast reactor model
Outline of SRT​’s microreactor model
SRT produces detailed output such as this fractional release plot, providing an overview of the transport and retention of isotopes within the fuel and compartment model.
Example SRT output showing the time-dependent radionuclide inventory in a reactor building compartment.