Hugo Cesar Rezende

Power calibration of the TRIGA mark I nuclear research reactor

This paper presents the results and methodology used to calibrate the thermal power of the TRIGA Mark I IPR-R1 Research Reactor at the Nuclear Technology Development Centre (CDTN), in Belo Horizonte, Brazil. The TRIGA Mark I is a pool type reactor, cooled by water natural convection. The method used in the calibration consisted in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in this primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The thermal losses from the primary loop were not evaluated since the inlet and outlet temperatures were measured just above the water surface of the reactor pool. The temperature of the water in the reactor pool as well as the reactor room temperature were set as close as possible to the soil temperature to minimize heat leakages. These leakages are mainly due to the conduction through the concrete and metal walls and also due to the evaporation and convection through the water surface of the reactor pool.

Performance of some operating parameters of a TRIGA research reactor in natural convective cooling

In order to study the safety aspects connected with the permanent increase of the maximum steady state power of the IPR-R1 TRIGA Reactor of the Nuclear Technology Development Centre (CDTN) at Belo Horizonte (Brazil), experimental measures were performed with the reactor working in power steps from ‘zero’ power up to 265 kW. The experiments were performed with the pool forced cooling system turned off. A number of parameters were measured in real-time such as fuel and water temperatures, radiation levels, reactivity and influence of cooling system. This paper summarises the behaviour of the operational parameters and presents some of the recent results obtained. Information on all aspects of reactor operation was displayed on the Data Acquisition System (DAS) shown the IPR-R1 real-time performance. The DAS was developed to monitor and record all operational parameters. Information displayed on the monitor was recorded on hard disk in a historical database.

Experimental determination of heat transfer coefficients in uranium zirconium hydride fuel rod

The heat generated by nuclear fission is transferred from fuel elements to the cooling system through the fuel-to-cladding gap and the cladding-to-coolant interfaces. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. A correlation for the gap conductance between the fuel and the cladding was also presented. As the reactor core power increases, the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the reactor core at power levels in excess of 265 kW.

Experimental Investigation and Computational Validation of Thermal Stratification in Piping Systems of PWR Reactors

One phase thermally stratified flow occurs in horizontal piping where two different layers of the same liquid flow separately without appreciable mixing due to the low velocities and difference in density (and temperature). This condition results in a varying temperature dis‐ tribution in the pipe wall and in an excessive differential expansion between the upper and lower parts of the pipe walls. This phenomenon can induce thermal fatigue in the piping system threatening its integrity. In some safety related piping systems of pressurized water reactors (PWR) plants, temperature differences of about 200 oC can be found in a narrow band around the hot and cold water interface. To assess potential piping damage due to thermal stratification, it is necessary to determine the transient temperature distributions in the pipe wall (Häfner, 2004) (Schuler and Herter, 2004).

Thermal behaviour of the IPR-R1 TRIGA nuclear reactor

Experimental and analytical studies have been performed at the Nuclear Technology Development Center ? CDTN (Belo Horizonte) to find out the temperature distribution in the IPR-R1 TRIGA Research Nuclear Reactor, as a function of power and position in the reactor core. The basic safety limit for the TRIGA reactor system is the fuel temperature, both in steady-state and pulsed mode operation. The time dependence of temperature will not be considered here, hence only the steady-state temperature profile will be studied. The experimental results for fuel and coolant temperatures in the reactor core, at different reactor power levels, have been compared with theoretical data and some results from other TRIGA reactors.