Natural Circulation Modeling in MTR Fuel Geometry in Research Reactors

In many research reactors the reactor can operate under either forced or natural convection modes. Under forced convection, the primary cooling system removes the heat generated in the reactor core through a heat exchanger to the secondary cooling system, which releases this thermal energy through the cooling tower to the atmosphere. Under natural convection operating mode, the generated nuclear power imply heats up the pool water, and is ultimately dissipated through the pool surface to the containment atmosphere. Hence, the large pool provides a heat sink for the energy generated within the core. The pool surface is open to the containment atmosphere, where thermal energy exchange via evaporation and convection heat transfer can occur. In case of electrical supply failure or mechanical / electrical failure in the pumps, a Loss of Flow Accident (LOFA) will occur. In this case natural circulation mechanism may be the only mechanism that can be use to remove the decay heat from the core. Modeling and modeling validation of natural circulation is an important issue in designing and commissioning of nuclear reactors. Most of the designers are using commercial codes like RELAP5, TRAC and THYDE-W to simulate the behavior of the core under natural circulation. Validation of those cods is done by comparing between different codes and comparing the calculated results with results from scaled experimental systems. In the last two decades, CFD calculations are also used for calculations and validation of the commercial codes. In case of research reactors and in case of Loss of Flow Accident, a simple analytical model can be use for preliminary and advanced calculations of the reactors. The purpose of this paper is to present and validate a simple model for thermohydraulic calculation of research reactor in natural circulation mode