RELAP5-3D Analysis of EBR-II Shutdown Heat Removal Test SHRT-17

Abstract

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. The aim of this paper is to present modeling of EBR-II reactor design using RELAP-3D, to show the results of the transient analysis of SHRT-17, and to discuss the results of application of the Fast Fourier Transform Based Method (FFTBM) to perform a quantitative accuracy evaluation of the model developed. 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). 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. Finally, the process of qualification of a system thermal-hydraulic code calculation was applied. It consists of three steps: 1) the geometrical fidelity of the nodalization, related with the evaluation and comparison of the geometrical data of the hardware respect to the estimated numerical values implemented in the nodalization; 2) the steady state level qualification, dealing with the capability of the nodalization to reproduce the steady state qualified conditions of the system; 3) the “on-transient” qualification, necessary to demonstrate the capability of the code and of the developed nodalization to reproduce the relevant thermal-hydraulic phenomena expected during the transient. The latter is a very complex step which foreseen different phases following our methodology of qualification (SCCRED, Standardized and Consolidated Calculated & Reference Experimental Database methodology). In the framework of the benchmark, the focus was only on the so called “Quantitative Accuracy Evaluation” that is performed by the FFTBM.