The PARET code, originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors, has now been modified to include a selection of flow instability, departure from nucleate boiling, single- and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model that estimates the voiding produced by subcooled boiling. This modified code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons have been made with the experimental results from the SPERT-I transients, and the agreement with the experimental data is generally quite good. The selection of proper correlations and properties for the range of interest in research reactors was essential to the accuracy of the results. The code has also been applied to the analysis of the International Atomic Energy Agency 10-MW benchmark cores for protected and unprotected transients. The code provides an accurate capability for the analysis of research reactor transients. This modified version of the PARET code is available through the National Energy Software Center.