Manmohan Pandey

Role of Fine Needle Aspiration Cytology in Head and Neck Lesions of Paediatric Age Group

CONTEXT Fine Needle Aspiration Cytology [FNAC] of the head and neck region is well accepted as a diagnostic procedure. Various studies in the context of FNAC in the head and neck region are available for the adult population, but only few studies are available for the paediatric age group. AIMS To study the role of fine needle aspiration cytology and its utility in paediatric head and neck lesions. SETTINGS AND DESIGN This was a hospital based, prospective study. METHOD AND MATERIALS Hundred cases of head and neck lesions of the paediatric age group [0-15 years] were studied for cytomorphology through fine needle aspiration cytology and the results were correlated with the histomorphology. RESULTS There was a male predominance in the case distribution among both the sexes in children [55%]. The head and neck lesions were most frequent in the age group of 10-15 years, followed by the age group of 5-10 years than the age group of 0-5 years. Lesions in the cervical lymph nodes constituted 81% of the head and neck lesions and 87% of the adequate smears, followed by those in the skin and subcutaneous tissues [3 cases (3.2%)], the thyroid [4 cases (4.3%)] and the salivary gland [1 case (1%)]. 88.17% cases of head and neck lesions in children were diagnosed as benign on their smears and 11.83% cases were diagnosed as malignant, of which 8 cases of malignant lesions were located in the cervical lymph nodes, 1 case was located in the thyroid and 2 cases of malignant lesions were located in the orbits. CONCLUSIONS FNAC is an important and a non-invasive, investigational tool in children for identifying and planning the medical management of inflammatory and infectious conditions. It helped us in indicating the diagnosis of the lesions in congenital or aquired malformations, cystic lesions and benign neoplastic lesions, in which surgical management were needed and we got confirmations on histological examinations. For the malignant lesions, FNAC was a more important investigation tool than an accurate investigation tool, which suggested about the lesions and guided us to do more advanced specific investigations for obtaining the diagnosis.

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

Improved Scaling Analysis for Heat Transfer in a Circular Tube With Various Supercritical Fluids Using Computational Fluid Dynamics Simulations

ABSTRACT The operating conditions of supercritical water cooler reactor (SCWR) are well above the critical point of water, so it is not possible to investigate its heat transfer aspects through laboratory experiments without industry-scale support. The most feasible alternative can be to scale-down the operating parameters by fluid-to-fluid scaling with a suitably chosen scaling fluid. However, it is impossible to incorporate all phenomenological factors of an intricate system like the SCWR through simple analytical scaling. This study demonstrates the limitation of fluid-to-fluid scaling in such situations and suggests the incorporation of computational fluid dynamics simulation as a subsequent step for better scaling. A scaling methodology from the published literature is adopted. Carbon dioxide and R134a have been considered as scaling fluids to identify the parameter ranges suitable for lab-scale simulation of the SCWR. A circular tube of 8 mm diameter and 1500 mm length is taken for simulation. A grid dependency test is done and the standard κ − ϵ turbulence model is selected. The developed computational model showed amicable agreement with existing experimental data. Analytically scaled-down parameters failed to simulate the axial and radial temperature profiles of the prototype. Increase in wall heat flux and reduction in mass flow rate are suggested as two possible options for achieving better profile matching. The modified values of scaled parameters with respect to a particular prototypical condition are reported. Profiles with CO2 as model fluid show better agreement with water as compared to R134a and hence this is recommended for use in lab experiments.

Numerical Investigation of Startup Instabilities in Parallel-Channel Natural Circulation Boiling Systems

The behaviour of a parallel-channel natural circulation boiling water reactor under a low-pressure low-power startup condition has been studied numerically (using RELAP5) and compared with its scaled model. The parallel-channel RELAP5 model is an extension of a single-channel model developed and validated with experimental results. Existence of in-phase and out-of-phase flashing instabilities in the parallel-channel systems is investigated through simulations under equal and unequal power boundary conditions in the channels. The effect of flow resistance on Type-I oscillations is explored. For nonidentical condition in the channels, the flow fluctuations in the parallel-channel systems are found to be out-of-phase.

Analysis of Coupled Neutronic-Thermohydraulic Instabilities in Supercritical Water-Cooled Reactor by Lumped Parameter Modeling

The possibility of instabilities in future nuclear reactors cooled by supercritical water is a matter of concern due to sharp changes in thermodynamic properties of coolant within the core. In the present work, a lumped parameter dynamic model of supercritical water-cooled reactor has been developed for analysis of coupled neutronic-thermohydraulic instabilities. The coolant channel is divided into two nodes with a moving boundary between them. The heater wall dynamics is described by a lumped parameter energy balance. Point neutron kinetics with one group of delayed neutrons has been used to model the power dynamics. Simple non-dimensional equations of state have been obtained for evaluating thermodynamic properties. Stability analysis has been done for various values of parameters such as the reactor power, coolant mass flow rate and inlet temperature. Stability maps have been plotted in the parameter planes. Dynamic simulations have been performed in the time domain to study the nature of operating transients. The stability analysis with neutronics is found to be more conservative. Transient simulations without neutronics indicate a supercritical Hopf bifurcation and the existence of a stable limit cycles in the unstable region. However, simulations with coupled neutronics indicate a subcritical Hopf bifurcation and the existence of unstable limit cycles in the stable region. Therefore, the analysis with neutronics is more conservative and shows that the system can be unstable for large perturbations, even if it is stable for small perturbations.Copyright

Experimental and Computational Studies on Effects of Scale-Up of Riser on Heat Transfer Chatacterisitcs of Circulating Fluidized Bed

In the present paper, CFD simulations using ANSYS-Fluent 14.5 were accomplished to study the effect of bed or sand inventory and effect of scale-up of riser on heat transfer characteristics like temperature and heat transfer coefficient. To accomplish the scale-up study, 3D CFD simulations were performed on the Circulating Fluidized Bed (CFB) risers of cross section 0.15 × 0.15 m, 0.30 × 0.30 m, and each of height 2.85 m. CFD simulations to predict heat transfer characteristics were accomplished under same operating conditions on heated portion (heater) of both risers. The walls of heater were maintained at the constant heat flux q″ = 1,000 (W/m2). Modeling and meshing were done using ProE and ANSYS ICEM CFD software, respectively. RNG k-e model was used for turbulence modeling. Eulerian model with Gidaspow phase interaction scheme was used to simulate the two phase flow (air + sand mixture flow). Computational (CFD simulation) data was compared with experimental data for the validation purpose. After validation, further simulations were conducted on riser of cross section 0.30 × 0.30 m. Based on scale-up study, empirical correlation has been developed to predict the heat transfer coefficient.

Numerical Investigations on Instabilities in Supercritical Water Flowing Through Heated Parallel Channels

The supercritical water-cooled reactor (SCWR) is a concept for an advanced reactor that operates at supercritical pressure and temperature. However, near the pseudocritical region the coolant properties are very sensitive to temperature, which raises concern about flow instabilities, whose analysis is very important to ensure safety of SCWR. Existence and nature of instabilities can be investigated numerically, using commercial system codes based on one-dimensional unsteady equations, suitable equations of state for thermo-physical properties of supercritical water, and suitable correlations for hydraulic resistance and heat transfer in supercritical fluids. In the present work, numerical investigation of stability characteristics of heated parallel channel system carrying supercritical water has been carried out using the RELAP5 code. The inlet and outlet pressures, the inlet temperature, and the heat flux were specified as the boundary conditions. The numerical model was validated with the help of experimental results reported in the literature. Transient numerical simulations were carried out to study the effect of various operating parameters (e.g., the mass flow rate, the inlet temperature, and the heat flux). The trends obtained were similar to those reported in the literature, albeit the asymmetry was less pronounced due to higher system pressure.© 2013 ASME

Study on Design of Scaled Down Test Facilities for Investigation of Instabilities in Supercritical Water Reactor

Supercritical Water Reactor (SCWR) proposes higher thermal efficiency and simpler plant design compared to modern Boiling Water Reactors. High pressure, temperature and power requirement in SCWR, however, escalates the cost of an experimental facility significantly. Present work, therefore, focuses on designing downscaled test facilities for stability analysis of SCWR. The facilities are conceptualized to model the European reference design of SCWR under both forced and natural circulation condition. R-134a is identified as the scaling fluids through fluid-to-fluid modeling, along with two others from literature. Similarity variables are obtained following two different approaches, starting from fundamental conservation equations. Dimensional and non-dimensional representations of important geometric, kinematic and dynamic parameters are evaluated and compared. Comparisons between two different approaches, as well as between forced and natural circulation have been presented for each scaling fluid.

CFD Investigation of Heat Transfer Deterioration in Supercritical Water Flowing Through Vertical Annular Channels

In order to enhance the efficiency of current light water reactors, the generation IV initiative has included the supercritical water reactor (SCWR) as one of the future designs. The rapid change in density in the vicinity of the pseudo-critical temperature leads to strong buoyancy effect at low flow rates and flow acceleration at high flow rates, both of which significantly influence heat transfer characteristics. Experimental investigation of such phenomena being very cumbersome and cost-intensive, numerical simulation using CFD tools is considered to be a useful option for providing better understanding of the heat transfer mechanisms in geometries and conditions typical of SCWR. The present work involves numerical analysis of the heat transfer deterioration (HTD) phenomenon in turbulent flow of supercritical water through a vertical annular channel. ANSYS-CFX 14.0 software was employed for the same. An annular fluid domain, with a heated inner wall and an insulated outer wall, was modeled and the flow was considered to be in the upward and downward directions. Grid independence study was conducted with structured mesh. The results were compared with those reported in the published literature. It is known that the HTD phenomenon causes a sudden rise in the wall temperature, and hence it is necessary to predict the effect of changes in operating and design parameters. Parametric study was done by varying pressure, inlet temperature, heat flux and mass flux. Annuli of different hydraulic diameters were also considered.© 2013 ASME

Numerical Computation of One-Dimensional Unsteady Two-Phase Flow Using HEM Model and IAPWS IF-97 Equations of State

Numerical simulation of transients in two-phase flow is crucial to simulate accident-like and aberrant conditions of nuclear reactors for safety analysis. A considerable number of such problems can be treated as one-dimensional with significant reduction in complexity without much loss in applicability. Most commercial thermal hydraulic codes are based on the two-fluid model, which solves balance equations for each phase and also accounts for thermodynamic non-equilibrium between the phases. However, the homogenous equilibrium mixture (HEM) model of two-phase flow can be employed to develop simple and efficient codes for transient simulations, using which extensive parametric studies can be carried out. In the present work, a code for numerical computation of unsteady one-dimensional two-phase flow has been developed and reactor transients have been simulated. The governing equations were obtained by the HEM model and were decoupled and approximated using the sectionalized compressible flow (SC) model and the momentum integral (MI) model. The equations of state used in the code are based on IAPWS Industrial Formulation-97. Pressure and heat flux transients for PWR and BWR were simulated with the code and compared with those reported in the literature. Further numerical simulations with the code were carried out to predict the transient response of nuclear reactors to various perturbations.Copyright