Accident analysis in research reactors

The incomplete understanding of the complex mechanisms connected with the interaction between thermal-hydraulic and neutron kinetics stili chailenges the design and the operation of nuclear reactors and imposes the adoption of conservatism in the evaluation of safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience suggests the revisiting of those areas and the identification of designloperation requirements that can be relaxed. So far, almost all of the safety analyses of research reactors have been performed using conservative computational tools such as channel codes but, nowadays, the appiication of Best-Estimate (BE) methods constitutes a rea! necessity. The global aim of the current work is an attempt to apply the best-estimate system thermalhydraulic code Relap5. For this purpose, the generic IAEA research reactor Benchmark problem is re-considered for proving the adequacy of the available computationai tools. Within the same framework, one of the most severe accident categories that may occur during a research reactor iifetime is aiso considered. This is related to a total and partiai blockage of the cooling charme! of a singie Fuel Assembly. Such event constitutes a stern scenario for this type of reactor since it may lead to local dryout and eventually to the bss of the fuei assembly integrity. The study constitutes the first step of a !arger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. To demonstrate the suitabi!ity of the technique, the bss of Shutdown Heat Removal accident in a MTR poo! type research reactor is ana!ysed. The accident occurs when the passive shutdown naturai convection coo!ing system is fai!ing for instance due to the rupture of an experimenta! beam tube. The accident will !ead to a partia! core uncovering. Although most of the research investigations in the world were performed for the analysis of the natura! air coo!ed research reactor core, it is demonstrated that there is a power density may exist above which partial submergence causes higher temperature than no submergence at all