Benchmark Analysis of EBR-II Protected Loss-of-Flow Transient

Coordinated Research Project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT) was established by International Atomic Energy Agency (IAEA). The objective of the project is to support and to improve validation of simulation tools and projects for Sodium-cooled Fast Reactors (SFR). The Experimental Breeder Reactor II (EBR-II) plant was a uranium metal-alloy-fuelled liquid-metal-cooled fast reactor designed and operated by Argonne National Laboratory (ANL) for the U.S. Department of Energy at the Argonne-West site. In the frame of this project, benchmark analysis of one of the EBR-II shutdown heat removal tests, protected loss-of-flow transient (SHRT-17), has been performed at the Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG) in Pisa, Italy. The aim of this paper is to present modeling of EBR-II reactor design using RELAP-3D, and to present results of the transient analysis of SHRT-17. Complete nodalization of the reactor was made from the beginning. Model is divided in primary side that contains core, pumps, reactor pool and, for this kind of reactor specific, Z pipe, and intermediate side that contains Intermediate Heat Exchanger (IHX). Core was modeled with 82 channels that represent all fuel assemblies, and 14 channels for reflector and blanket assemblies. After achievement of acceptable steady-state results, transient analysis was performed. Starting from full power and flow, both the primary loop and intermediate loop coolant pumps were simultaneously tripped and the reactor was scrammed to simulate a protected loss-of-flow accident. In addition, the primary system auxiliary coolant pump, that normally had an emergency battery power supply, was turned off. Despite early rise of the temperature in the reactor, the natural circulation characteristics managed to keep it at acceptable levels and cooled the reactor down safely at decay heat power levels. Thermal-hydraulics characteristics and plant behavior was focused on prediction of natural convection cooling by evaluating the reactor core flow and temperatures and their comparison with experimental data that were provided by ANL