P. K. Vijayan

Steady-State Behavior of Natural Circulation Loops Operating With Supercritical Fluids for Open and Closed Loop Boundary Conditions

Supercritical water (SCW) exhibits excellent heat transfer characteristics and a high volumetric expansion coefficient (hence high mass flow rates in natural circulation systems) near the critical temperature. SCW is being considered as a coolant in some advanced nuclear reactor designs on account of its potential to offer high thermal efficiency, compact size, and elimination of steam generator, separator, and dryer, making it economically competitive. The elimination of phase change results in elimination of the critical heat flux phenomenon. Cooling a reactor at full power with natural instead of forced circulation is generally considered an enhancement of passive safety. In view of this, it is essential to study natural circulation behavior at supercritical conditions. Carbon dioxide can be considered to be a good simulant of water for natural circulation at supercritical conditions, since the density and viscosity variation of carbon dioxide follows a curve parallel to that of water at supercritical conditions. Hence, experiments were conducted in a closed supercritical pressure natural circulation loop (SPNCL) with supercritical carbon dioxide as working fluid. A nonlinear stability analysis code (NOLSTA) has been developed to carry out steady-state and stability analysis of open and closed loop natural circulation at supercritical conditions. The code has been validated for steady-state predictions with experimental data available in open literature and experiments conducted in SPNCL.

Simulation of natural circulation in a rectangular loop using CFD code PHOENICS

Abstract Single phase natural circulation in a rectangular loop is simulated using the PHOENICS code, a general purpose Computational Fluid Dynamics (CFD) code. The rectangular loop, having different operating power levels, has been modeled with the help of the Multiple Block Fine Grid Embedment (MBFGE) technique. The Co-located Co-variant Method (CCM) is used to simulate this loop in PHOENICS. Extensive experimental and CFD studies have been conducted on single phase natural circulation in a rectangular loop. The paper presents the results of three-dimensional CFD analysis for the prediction of steady state behavior in a rectangular loop and its comparison with experimental data. The results of code prediction and readily available experimental data show good agreement.

Numerical investigation of heat transfer in the vertical annulus between pressure tube and calandria tube of the advanced water cooled reactor

Abstract In advanced water cooled reactors, an annular gap exists between pressure tube and calandria tube. The gap is closed from top but is open from bottom. Due to differential temperature between pressure tube and calandria tube, air flow is induced by natural convection. This leads to heat transfer from pressure tube to calandria tube. The quantification of the heat transfer between pressure tube and calandria tube is numerically carried out with the help of the CFD code PHOENICS. Validation of the CFD code with experimental results and some established computational work from the literature has been done in order to verify the accuracy of the code. The natural convection phenomenon in the annular gap is then simulated. The velocity and temperature fields obtained from the CFD simulation are used to compute local and average heat transfer coefficients. Heat transfer coefficients for various pressure tube temperatures are computed. The effect of water on the heat transfer in the annular gas is also studied.

Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code

Abstract Three dimensional CFD simulations of the moderator flow in the calandria vessel of a heavy water reactor are performed using the PHOENICS CFD code. The model includes the entire calandria vessel consisting of three shells, calandria tubes and inlet and outlet nozzle openings. The computational model prepared in PHOENICS consists of (a) standard k-∊ turbulence model, (b) PARSOL technique for handling curved objects in cartesian grids and (c) Boussinesq formulation for handling variable density flows. PHOENICS is validated by applying it to three different flow cases. The flow pattern in the calandria vessel under normal operating conditions is obtained through simulation. The effect of the presence of calandria tubes and heat generation on moderator flow pattern is studied. The simulation is also performed for various heat loads and moderator mass flow rates. The maximum temperature achieved by the moderator flow under various heat loads and moderator mass flow rates is obtained.

Natural circulation in a rectangular loop with vertical heater below vertical cooler

Abstract Many upcoming new generation reactors employ natural circulation for heat transfer in normal mode of operation. Natural circulation systems are simpler and safer than their forced circulation counterparts. However, these systems are prone to flow instability which are undesirable due to several reasons. In the present work, a rectangular glass loop, wherein cooler is just above the heater, has been considered for experimental and numerical investigation at atmospheric pressure. Heat addition from room conditions has been studied to understand the natural circulation loop dynamics, checking the possibility of occurrence of instability with the new orientation of the heater and the cooler. Experiments were performed at different power levels and coolant flow rates. CFD analyses were performed for all the cases investigated experimentally using the commercial CFD code ANSYS FLUENT 14.0. No instability was observed during the experiments and none during the simulations done for the duration of the experiments.

Experimental studies in a single-phase parallel channel natural circulation system: preliminary results

Abstract Natural circulation systems find extensive applications in industrial engineering systems. One of the applications is in nuclear reactor where the decay heat is removed by natural circulation of the fluid under off-normal conditions. The upcoming reactor designs make use of natural circulation in order to remove the heat from core under normal operating conditions also. These reactors employ multiple vertical fuel channels with provision of on-power refueling/defueling. Natural circulation systems are relatively simple, safe and reliable when compared to forced circulation systems. However, natural circulation systems are prone to encounter flow instabilities which are highly undesirable for various reasons. Presence of parallel channels under natural circulation makes the system more complicated. To examine the behavior of parallel channel system, studies were carried out for single-phase natural circulation flow in a multiple vertical channel system. The objective of the present work is to study the flow behavior of the parallel heated channel system under natural circulation for different operating conditions. Steady state and transient studies have been carried out in a parallel channel natural circulation system with three heated channels. The paper brings out the details of the system considered, different cases analyzed and preliminary results of studies carried out on a single-phase parallel channel system.

Experimental investigations on control of flow instability in single-phase natural circulation loop

Abstract Natural circulation systems offer simplicity, enhanced safety and reliability and thus are advantageous over their forced circulation counterpart. However, natural circulation is susceptible to flow instabilities. These instabilities are undesirable for various reasons. Literature suggests the use of orifices at the inlet to suppress instability. However, orificing introduces pressure drop which limits the flow rate and hence the heat carrying capacity of the fluid. In the present study, investigations have been carried out with different mechanical gadgets to control the natural circulation flow instabilities. Extensive experiments have been carried out in a single-phase rectangular natural circulation loop to study the effect of these mechanical gadgets on instability. The paper brings out the results of experimental investigations carried out on the role of mechanical gadgets in controlling instability.

Assessment of the look-up table using the tubular and bundle CHF data and modification of the bundle correction factor

Abstract The Critical Heat Flux (CHF) is an important parameter, which limits the thermal hydraulic performance of the nuclear fuel bundle. The tools available for the prediction of the CHF are empirical in nature and are valid for their experimental range only. However, the recently developed Look-Up Table (LUT) approach has emerged to be a promising tool for predicting CHF in a tubular geometry over a wide range of parameters, which can be extended to the rod bundle geometry considering correction factors for the rod bundle effects. The error statistics of the present assessment confirms the values provided with the LUT for the HBM approach for the tubular application. However, the error statistics by DSM (not provided with the LUT development) is found to be quite different from that of the HBM. It is found that CHF in the rod bundle can also be predicted with a good accuracy using the Heat Balance Method (HBM). The proposed correction factor is found to improve the prediction accuracy of the LUT for the rod bundle application. This paper deals with the assessment of the CHF prediction by LUT for the tubular and bundle geometry and evaluation of the correction factor for rod bundle at the normal operating pressure of BWR (70 bar) using the experimental data base.