Evaluation of RELAP5/MOD3 behavior against loss of flow experimental results from two research reactor facilities

Benchmarking COMSOL Multiphysics Single-Subchannel Thermal-Hydraulic Analysis of a TRIGA Reactor with RELAP5 Results and Experimental Data

COMSOL Multiphysics has been used to conduct thermal-hydraulic analysis in multiple nuclear applications. The aim of this study is to benchmark the prediction accuracy of COMSOL Multiphysics in performing thermal-hydraulic analysis of TRIGA (Training, Research, Isotopes, General Atomics) reactors such as the Geological Survey TRIGA Reactor (GSTR) by comparing its predictions with RELAP5 (a widely used code in nuclear thermal-hydraulic analysis) results and experimental data. The GSTR type is Mark I with a full thermal power of 1 MW, and it resides at the Denver Federal Center (DFC) in Colorado. The numerical investigation of the present work is carried out by developing single-subchannel thermal-hydraulic models of the GSTR utilizing RELAP5 and COMSOL codes. The models estimate the temperatures (fuel, outer clad, and coolant) and water flow patterns in the core as well as fuel element powers at which void starts to form within the coolant subchannels. Then, these models’ predictions are quantitatively evaluated and compared with the measured data.

Simulation and analysis of the Loss of Flow Accident (LOFA) scenarios for an open pool type research reactor by using the RELAP5/MOD3.2 code

In this paper, a 5 MW open pool type research reactor is simulated and analyzed to check and pass a complete and practical safety assessment against various LOFA scenarios. Generally, LOFA may occur in a research reactor when the reactor primary pump or the safety flapper valve fail. The simulation includes the study of a downward core cooling system circulated by the gravity driven force, the performance and failure effects of the safety flapper valve, the flow reversal mode and transition, the natural convection mode, different possible types of LOFA, and impacts of the surrounding empty boxes. First of all, the code nodalization is successfully benchmarked against available experimental data of the reactor operating parameters. Then, all possible DBAs in this field of study are simulated and discussed in detail. Results are completely satisfactory to simulate and analyze a pool-type research reactor in response to LOFA using the RELAP5 code. Furthermore, transients including the natural convection mode and even flow reversal mode are following naturally without any oscillation or source of errors. Finally, TRR is completely safe against the DBA type of LOFA scenarios.