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COB43 REPRESENTAÇÃO DOS SISTEMAS AUTOMÁTICOS DE CONTROLE E DE PROTEÇÃO DE ANGRA I / SIMULATION OF THE ANGRA I CONTROL AND PROTECTION SYSTEMS

Ronaldo C. Borges & Anibal N. Gebrim

Comissão Nacional de Energia Nuclear (CNEN), Coordenação de Reatores

Rua General Severiano 90 - Sala 419/A, 22294-900, Rio de Janeiro, RJ, Brasil, E-mail: ronaldo@cnen.gov.br e anibal@cnen.gov.br

The following automatic control systems for Angra I Nuclear Power Plant (NPP): rod control system, steam generator level control system, feedwater bypass valve control system, pressurizer pressure control system and pressurizer level control system and also the following protection system: overtemperature, overpower, pressurizer low pressure, pressurizer high pressure, pressurizer high water level, low primary coolant flow and low-low steam generator water level were simulated with the RELAP5/Mod2 code. The paper shows the results of three transients: a) a ten percent step load change, b) a five percent per minute ramp load change, and c) pump coastdown. Satisfactory quantitative and qualitative results were obtained when comparing these results with the "setpoint study" and FSAR results.

Keywords: Control System, Protection System, Nuclear Power Plant, Angra I, RELAP5/Mod2 Code

Sistema de Controle, Sistema de Proteção, Usina Nuclear, Código RELAP5/Mod2

COB75 COMPORTAMENTO TERMOIDRÁULICO DE VARETAS AQUECIDAS ELETRICAMENTE DURANTE TRANSITÓRIO DE FLUXO CRÍTICO DE CALOR/ THERMALHYDRAULIC BEHAVIOR OF ELECTRICALLY HEATED ROD DURING A CRITICAL HEAT FLUX TRANSIENT

Rita de Cássia Fernandes de Lima & Pedro Carajilescov*

Departamento de Engenharia Mecânica, Universidade Federal de Pernambuco - Av. Acad. Hélio Ramos, s/n, 50740-530 Recife, PE, Brasil

E-mail:ritalima@npd.ufpe.br

*Departamento de Engenharia Mecânica, Universidade Federal Fluminense - Rua Passo da Pátria,156, 24210-240 Niterói, RJ, Brasil

E-mail: pedroc@caa.uff.br

In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, a theoretical analysis of the drying and rewetting front propagation is performed during a critical heat flux experiment, starting with the application of a slope of electrical power from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies done with several values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one.

Keywords: Critical heat flux, rewetting front, drying front, thermalhydraulics, numerical simulation. / Fluxo crítico de calor, frente de remolhamento, frente de secamento, termoidráulica, simulação numérica.