D. F. Delmastro

Centro Atomico Bariloche,

San Carlos de Bariloche, Argentina

Abstract. CAREM is an innovative reactor with an integrated self-pressurized primary system, developed by Argentina. The primary system coolant circulation is of natural circulation type and several passive safety systems are included. The thermal-hydraulic behavior of CAREM reactor was study using generic numerical codes. Several transients and accidental situations were analyzed. A High Pressure Natural Convection Loop was constructed and operated to produce data in order to verify the thermal hydraulic tools used to design CAREM reactor, mainly its dynamical response. This is accomplished by the validation of the calculation procedures and codes for the rig working in states that are very close to the operating states of CAREM reactor. Several dynamical experiments were performed and new ones are planned. The data obtained is being used to test our numerical procedures and codes. In this paper an overview of the thermal-hydraulic aspects of CAREM reactor is presented. The analytical dynamical studies and experimental facility, studies and results are briefly presented.

1. INTRODUCTION

The CAREM nuclear power plant design is based on a light water integrated reactor. The whole high-energy primary system, core, steam generators, primary coolant and steam dome, is contained inside a single pressure vessel.

The flow rate in the reactor primary systems is achieved by natural circulation. Figure 1 shows a diagram of the natural circulation of the coolant in the primary system. Water enters the core from the lower plenum. After been heated the coolant exits the core and flows up through the riser to the upper dome. In the upper part, water leaves the riser through lateral windows to the external region. Then it flows down through modular steam generators, decreasing it enthalpy. Finally, the coolant exits the steam generators and flows down through the down-comer to the lower plenum, closing the circuit. The driving forces obtained by the differences in the density along the circuit are balanced by the friction and form losses, producing the adequate flow rate in the core in order to have the sufficient thermal margin to critical phenomena. Reactor coolant natural convection is produced by the location of the steam generators above the core. Coolant acts also as moderator.

Self-pressurisation of the primary system in the steam dome is the result of the liquid-vapour equilibrium. The large volume of the integral pressuriser also contributes to the damping of eventual pressure perturbations. Due to self-pressurisation, bulk temperature at core outlet corresponds to saturation temperature at primary pressure. Heaters and sprinkles typical of conventional PWR’s are thus eliminated.