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.
In the framework of a BEPU (Best Estimate Plus Uncertainty) approach within the licensing process of a nuclear power plant, the need to extend the resources of nuclear system thermal-hydraulics codes, such as RELAP5-3D, arises to allow more detailed simulations of the complex 3D reality of Nuclear Power Plants (NPPs), either under normal steady-state or during various accident scenarios. Currently, it is not possible to achieve the same degree of detail for a whole nuclear system when it is simulated with RELAP5-3D and this is due to the inherent limitations in the number of components and volumes to be used for the analysis. For this reason, it is of extreme interest the use of tools for codes coupling that enable the use of different codes for the simulation of different portions of a system in a unified analysis. In this paper the attention will be focused on the decomposition of the thermal-hydraulic domain of a system into subsystems to be simulated by different instances of the same code (e.g. RELAP5-3D) coupled together by means of PVMEXEC program and parallel virtual machine (PVM) technology. Explicit and semi-implicit solution algorithms were used for the analyses. Among the analyzed cases, the following will be discussed in detail with the aim to provide additional guidelines for the use of the PVMEXEC tool: (i) the Edward’s pipe blowdown test, (ii) a simplified countercurrent heat exchanger, (iii) different hydraulics and heat structure coupling schemes for a shell-tube heat exchanger and (iv) a three-task coupled model of a simplified BWR model.
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