Kannan N. Iyer

CFD analysis of single-phase flows inside helically coiled tubes

It has been well established that heat transfer in a helical coil is higher than that in a corresponding straight pipe. However, the detailed characteristics of fluid flow and heat transfer inside helical coil is not available from the present literature. This paper brings out clearly the variation of local Nusselt number along the length and circumference at the wall of a helical pipe. Movement of fluid particles in a helical pipe has been traced. CFD simulations are carried out for vertically oriented helical coils by varying coil parameters such as (i) pitch circle diameter, (ii) tube pitch and (iii) pipe diameter and their influence on heat transfer has been studied. After establishing influence of these parameters, correlations for prediction of Nusselt number has been developed. A correlation to predict the local values of Nusselt number as a function of angular location of the point is also presented.

Scaling of Natural Circulation Boiling Systems

This paper lays out the procedure for arriving at the dimensions of a model facility to simulate a pressure tube type reactor. The Advanced Heavy Water Reactor, whose design is being evolved in the Indian scenario, is used as a basis for the evolution of the model facility. The non-dimensional groups that need to be preserved are identified and the design is evolved by satisfying these non-dimensional groups. The inevitable distortions that get introduced are discussed and a suitable compensation procedure evolved. Finally, the evolved model is shown to satisfy both steady state and characteristic equation similarity.© 2003 ASME

Experimental and Numerical Investigation on a Two-Phase Natural Circulation Test Facility

A geometrically similar model of the proposed Advanced Heavy Water Reactor being designed and built by Bhabha Atomic Research Centre has been conceived and built at Indian Institute of Technology, Bombay. The basic objective of setting up such a facility and conducting experimental studies was to simulate the thermal-hydraulic behavior during the startup and also during part or full power operation. Owing to hydrodynamic instabilities observed in natural circulation system, it is necessary to know the stability boundaries. In this paper, the details of the experimental facility, experimental studies, and numerical simulations carried out using RELAP5/MOD 3.2 are discussed. Type I and type II stability boundaries and stable two-phase operation zones have been experimentally obtained at various pressures and inlet subcooling and the results are correlated using appropriate nondimensional numbers. Further, the experimental facility is modeled using the thermal-hydraulic code RELAP5/MOD 3.2. The single-phase pressure drops measured in the loop are used to estimate the appropriate loss coefficients used in the model. The model is then used to predict stability boundaries, and the predictions are compared with the experimental findings. The details of the experimental model and simulation results are presented. The capability of RELAP5/MOD 3.2 to predict the flow oscillations during instabilities is discussed.

Simulation for Startup Transients in a Natural Circulation Boiling Loop

Startup of natural circulation boiling water reactors is of current interest due to the transients that occur at low pressure and low power conditions. Numerical simulation can be a useful tool for studying startup transients and for devising appropriate startup procedure. In the present work, a numerical model of an experimental test facility has been developed with RELAP5/MOD3.4. This model has been used to devise a startup procedure, which has been successfully implemented in the test facility. Numerical simulations for the startup transients have been done with the RELAP5 model and results have been compared with experimental findings. The nature of oscillations predicted by numerical simulations is similar to that observed in the experiment. This study demonstrated one possible method for developing startup procedures for natural circulation boiling water reactors.Copyright

Comparison of Lumped Parameter and CFD Code Predictions: Sump Evaporation Phenomena

In nuclear reactor containment, Sump, Calandria Vault and Calandria Vessel contain large amounts of water. Condensation on walls and containment spray system actuation also results in accumulation of water in the containment sump during the accident conditions. Evaporation of water takes place during the accident conditions and needs to be accounted in the hydrogen distribution analysis. Lump parameter codes such as ASTEC have built-in models for sump water evaporation. However, CFD codes are being increasingly used for containment hydrogen distribution studies, development of a sump water evaporation model for multi-dimensional calculations is required. The sump model is implemented through mass and energy balance using two different approaches. The main focus of the paper is on the simulation of sump evaporation experiment conducted in TOSQAN facility using the lumped parameter code and its comparison with the CFD results and the available experimental data.

Hydrogen Distribution in a Nuclear Reactor Containment

The management of hydrogen in nuclear reactor containment after LOCA is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. Accurate turbulence modelling is important to predict the concentration distribution correctly. The turbulence models which were most commonly cited in the literature for modelling buoyancy driven flows were assessed for their suitability and it was found that the buoyancy modified Standard k-e model is adequate for the purpose by comparing with some experimental data available in the literature. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian nuclear reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. Results of all these cases have been presented in this paper.Copyright