[en] This paper systematically develops the scaling laws that have to be satisfied between a model and the prototype for their identical non-dimensional steady state and transient behaviour. From the rules derived, the dimensions of a scaled model for an assumed reactor configuration are worked out. (author)

[en] A liquid helium cryostat has been fabricated to be coupled to a Instron Model 1195 Universal Testing Machine for mechanical testing of materials at low temperatures. The sample is held between two grip assemblies, the top one being connected to the load cell by a long pull-rod. The bottom grip is connected to a tube, co-axial to the pull-rod, which in turn is attached to the moving cross-head. Studies with this set-up give information about the yield characteristics at room temperature, 77 K and at 4.2 K. With a manostat, the temperature ranges can be extended to 64-80 K and 1-5 K. The effect of magnetic field on yield characteristics, especially due to the viscous drag of dislocations by electron cloud, can be studied by a 10 kG superconducting magnet around the sample. The cryostat has been tested to the load capacity which is 5 KN. Results of tests done on annealed copper with this set-up are in agreement with the literature. Temperature stability at liquid nitrogen temperature makes it possible to do stress relaxation experiments for a duration of 15 minutes. The cryostat has a liquid helium consumption of about 3 litres per test. (author)

[en] Full text of publication follows: Hydrogen release and distribution in nuclear power plant containment is an important safety issue. Selection of a proper turbulence model is important for accurate estimation of the mixing process. The selection of turbulence model is dictated by the best compromise between accuracy and computational efforts. For this, three different turbulence models, viz. Standard k-ε, RNG k-ε and Reynolds Stress Model, based on Reynolds averaged Navier Stokes equations (RANS) approach, were used. The computations were done using the CFD code FLUENT, which is based on the control volume methodology. The computational results were compared with the experimental results of HYMIS test facility, where helium was used to simulate hydrogen. The processes of helium plume rise, multiple plume merging, distribution and mixing were studied. Based on these computations, a simple analytical/empirical zone based model was formulated for the same problem, which predicted the helium concentration reasonably accurately and quickly. (authors)

[en] Full text of publication follows: The phenomenon of carryover, i.e. entrainment of liquid along with separated steam is observed in all the steam separators. Due to the risks, such as turbine blade erosion and radioactivity leakage, associated with it, it is desired to have an estimate of the carryover value. This is all the more important when the separation is only under the influence of gravity, as in case of the proposed Advanced Heavy Water Reactor in India. From the previous experimental and theoretical investigations, an empirical model for prediction of liquid entrainment was identified. To validate this model, experiments were conducted in an air water facility and some discrepancies were observed between the experimental and predicted results. This was attributed to the difference in the diameters of the riser and the separator, which was not accounted for in the previous investigations. In order to ascertain this, further experiments were conducted with riser diameter to be same as the separator diameter, the results of which agreed well with the identified empirical model. To resolve the effects of diameter ratio, geometry, liquid and gas flux (over the interface and in the riser), and the height above the interface where entrainment is measured, a set of experiments were performed. Based on the above comprehensive data, an empirical correlation is proposed to predict the entrainment, which has a wider range of applicability than the currently available models. (authors)

[en] A compact ²³³U fuelled neutron source reactor is under construction at the Reactor Research Centre, Kalpakkam, India, to be used primarily for Neutron Radiography of fuel pins, subassemblies and other active components of the Fast Breeder Test Reactor. This BeO reflected source reactor uses approx. 550 gm of ²³³U as fuel in the form of U-Al alloy plates, and is moderated and natural convection cooled by light water. The reactor is designed to operate at 30 kW nominal power but can be uprated to 100 kW. The neutron flux in the core is as high as approx. 10¹² n/cm²s at 30 kW due to the small core size resulting in fluxes of approx. 3 x 10⁶ to 10⁷ n/cm²s at the radiography site. This compact source reactor is located underneath a hot cell where active fuel elements are to be visually examined. A leak tight tube extending down from the cell above permits active objects to be lowered for Neutron Radiography. Sequential pictures will be taken using cassettes driven on a chain conveyor system. Besides the beam for active sample radiography an additional beam would be available for non active sample radiography. The beam tubes are to be provided with standard collimator, aperture control and beam shutter arrangements. (Auth.)

[en] This paper presents a self-similar solution of the coupled problem of magneto-hydrodynamic free convection flow of an electrically conducting fluid arising from melting of a semi-infinite solid substrate. At steady state, buoyancy induced free convection of the electrically conducting fluid is influenced by the Lorentz force. A set of governing PDEs is developed for a two dimensional boundary layer problem including phase change which is simplified to a set of ODEs using a similarity transformation and are solved iteratively using an implicit Keller-box method. An asymptotic analytical solution for melting and heat transport rates is also presented for the case of small Prandtl numbers. The effect of each of the three characteristic parameters, viz., the Prandtl number, the melting parameter and the Lykoudis number on the similarity velocity and temperature profiles in the boundary layer over melting substrate is studied. It is observed that increasing the Lykoudis number or decreasing the Prandtl number lowers the melting rate and heat transfer at the substrate-melt interface. The use of magnetic field in controlling the free convection heat transfer, the melting rate and the thickness of the velocity and thermal boundary layers over melting substrate is elucidated and discussed. (authors)

No abstract available

[en] Experimental studies have shown that there is heat transfer enhancement (HTE) for supercritical fluid near the pseudocritical temperature at relatively low heat flux to mass flux ratios. At very high values of heat flux, a peak in wall temperature appears due heat transfer deterioration (HTD). In the present research work R22 has been selected as working fluid as the simulant fluid for water. A supercritical Freon test facility is then designed and built. Two vertical tubular test sections of ID equal to 6 mm and 13.5 mm are employed. Experiments with vertically upward flow at 55 bar system pressure were carried out. Thermal camera is used to obtain wall temperatures at distances about 1-1.5 mm apart. Experiments were conducted with water before using R22 to validate experimental procedure. Initial experiments with R22 were conducted to demonstrate the reduction in peak heat transfer enhancement with increase in heat flux. Experiments are then conducted at several heat and mass flux values and inlet temperature. It is observed from experimental results that onset of HTD occurs when q/G more than 0.056 to 0.072 (kJ/kg). When the inlet temperature is lowered, the onset of HTD appears at relatively high q/G. The bulk fluid enthalpy and temperature at which onset of HTD appears also reduces when the inlet temperature was decreased. At the lower inlet temperature, two peaks in wall temperature were observed in the experimental results

[en] Full text: Hydrogen mixing/distribution in nuclear reactor containment atmosphere is one of the important safety issues under accident conditions. It is therefore necessary to predict the hydrogen distribution accurately. Computational Fluid Dynamics (CFD) codes are increasingly used for this purpose. However, the commercially available CFD codes do not have the condensation models and one has to incorporate these models before applying these codes for this purpose as condensation of steam affects the distribution of the hydrogen. Further, during accident conditions, release of hydrogen into the containment may form combustible mixtures and which can pose a threat to the containment integrity. Re-combiners are placed in the containment to mitigate such risk by removing hydrogen from the containment atmosphere by catalytic reaction. Hence, re-combiner models also need to be implemented in the CFD codes to predict the hydrogen distribution. Finally, the turbulence models play a key role in dictating the convective diffusive behavior of hydrogen transport. The present paper systematically describes the selection of the condensation and re-combiner models to be implemented based on the model capabilities, applicability and their validation aspects. In addition, it also brings out the most competitive turbulence model in terms of accuracy and computational speed

[en] The heat transfer behaviour of a supercritical fluid is an important input for the performance evaluation of SuperCritical Water-cooled Reactors (SCWR) of Generation IV nuclear power plants. SCWRs have very high overall thermal efficiency of about 45-50% because their operating pressures and temperatures are high. Accident analyses for licensing are carried out using system thermal hydraulics codes in which well established and validated correlations for Heat Transfer Coefficient (HTC) have to be built in. Experimental studies have shown that there is Heat Transfer Enhancement (HTE) for supercritical fluid near the pseudocritical temperature at relatively low heat flux to mass flux ratios. The peak HTC decreases as this ratio increases. At very high values of heat flux, a peak in wall temperature appears due to Heat Transfer Deterioration (HTD). This phenomenon is important as the nuclear fuel clad may fail at high temperatures induced by HTD and result in release of radioactive nuclides into the coolant streams. It is also important for the sizing of core. Several investigators have carried out experiments using water, CO2 and R22 etc. due to similarity in their thermophysical properties, and it was observed that the bulk fluid enthalpy at which peak wall temperature appears is different at different test specifications (heat flux, mass flux and different inlet temperature). Effects of inlet temperature are found to be significant for HTD but results reported in literature have not reported this effect in detail. Experimental wall temperatures available in open literature were measured by thermocouples positioned at regular intervals. However, since the HTD is shown to result in very steep temperature changes, highly local temperature measurements are desirable. HTC correlations are available which are able to predict HTE satisfactorily but HTD predictions from available correlations are poor and therefore, better correlations are required to predict HTD.