Suscripción a Biblioteca: Guest
Proceedings of the 24th National and 2nd International ISHMT-ASTFE Heat and Mass Transfer Conference (IHMTC-2017)

ISBN En Línea: 978-1-56700-478-6


DOI: 10.1615/IHMTC-2017.820
pages 589-592

A. K. Vishnoi
Reactor Engineering Division, Bhabha Atomic Research Centre Mumbai, Maharashtra, India

Arnab Dasgupta
Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai-400085, India

D. K. Chandraker
Reactor Engineering Division, Bhabha Atomic Research Centre (BARC), Mumbai-400085, India

Arun Kumar Nayak
Homi Bhabha National Institute, Anushakti Nagar, Mumbai 400094, Maharashtra, India; Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai-400085, India

D. N. Badodkar
Reactor Design and Development Group, BARC, Mumbai


Thermal hydraulics plays a vital role in the safety evaluation of the nuclear reactor. The safety analysis envisages prediction of the thermal hydraulic parameters under normal operating conditions and anticipated transients as well. The licensing criteria demands that the critical parameters like the fuel temperature, clad temperature and critical heat flux should be well within the prescribed limit for the safe operation of the reactor. To meet this requirement, various computer codes are developed to analyze the normal operating conditions and the accidental scenarios. However, these codes need to be validated to minimize the prediction uncertainties of various models incorporated in the code. Thus validation of the codes demands generation of experimental data under the conditions of interest. Since the thermal hydraulic experiments in the actual operating nuclear fuel bundle is not possible due to radiation hazard and limitation on the instrumentations of the fuel bundle, in-pile experiments are generally ruled out. Also, these data needs to be made available before the licensing of the rector at the commissioning stage. Thus, in view of the safety requirement and limitation of the in-pile experiments, the research community generally relies on the out-of-pile experiments. In the out-of-pile experiments the nuclear fuel bundle is simulated using the electrical heating.

For typical bundles used in reactors, the power requirement for such experiments is high. The availability of such high power and complexity of associated process systems makes such experiments very expensive. In view of this, a new concept has been proposed for experiments which use symmetry sector of the bundle. This reduces the cost and time involved for experiments significantly. This paper deals with challenges in fabrication of such a symmetry sector Fuel Cluster Simulator (FCS) and it's equivalence with the full bundle.
The paper also presents typical transient during Critical Heat Flux (CHF) experiments.