R.K. Sinha

Linear and Nonlinear Stability Analysis of a Supercritical Natural Circulation Loop

Supercritical water (SCW) has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water cooled reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN has been developed employing supercritical water properties to carry out the steady-state and linear stability analysis of a SCW natural circulation loop (SCWNCL). The conservation equations of mass, momentum, and energy have been linearized by imposing small perturbation in flow rate, enthalpy, pressure, and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been benchmarked against published results. Then the code has been extensively used for studying the effect of diameter, heater inlet temperature, and pressure on steady-state and stability behavior of a SCWNCL. A separate computer code, NOLSTA, has been developed, which investigates stability characteristics of supercritical natural circulation loop using nonlinear analysis. The conservation equations of mass, momentum, and energy in transient form were solved numerically using finite volume method. The stable, unstable, and neutrally stable points were identified by examining the amplitude of flow and temperature oscillations with time for a given set of operating conditions. The stability behavior of loop, predicted using nonlinear analysis has been compared with that obtained from linear analysis. The results show that the stability maps obtained by the two methods agree qualitatively. The present paper describes the linear and nonlinear stability analysis models and the results obtained in detail.

Reliability assessment of passive isolation condenser system of AHWR using APSRA methodology

In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) is used for evaluation of reliability of passive isolation condenser system of the Indian Advanced Heavy Water Reactor (AHWR). As per the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to perform the design basis function. The methodology first determines the operational characteristics of the system and the failure conditions based on a predetermined failure criterion. The parameters that could degrade the system performance are identified and considered for analysis. Different modes of failure and their cause are identified. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the isolation condenser system performance. Once the failure surface of the system is predicted, the causes of failure are examined through root diagnosis, which occur mainly due to failure of mechanical components. Reliability of the system is evaluated through a classical PSA treatment based on the failure probability of the components using generic data.

Study on the flow-pattern-transition instability in a natural circulation heavy water moderated boiling light water cooled reactor

A mathematical model has been developed to study the flow pattern transition instability which may occur in a boiling two-phase system. The model considers flow pattern transition criteria for vertical upward and horizontal flow in pipes to identify the flow pattern transition and flow pattern specific pressure drop models. It also considers the drift flux model to estimate the void fraction in the two-phase region. The model has been applied to predict the flow pattern transition instability in a natural circulation heavy water moderated boiling light water cooled reactor. It is found that the instability characteristics is similar to that of the Ledinegg-type instability. However, the number of multiple steady states for a given operating power can be much larger in the flow pattern transition instability as compared to that of the Ledinegg-type instability. Stability maps were plotted and compared for both the flow pattern transition instability and that of the Ledinegg-type instability. The influence of various geometric and operating parameters on this instability were investigated.

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.

Linear and Non-Linear Stability Analysis of a Supercritical Natural Circulation Loop

Supercritical water has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN has been developed employing supercritical water properties to carry out the steady state and linear stability analysis of a SCW natural circulation loop. The conservation equations of mass, momentum and energy have been linearised by imposing small perturbation in flow rate, enthalpy, pressure and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been benchmarked against published results. Then the code has been extensively used for studying the effect of diameter, heater inlet temperature and pressure on steady state and stability behavior of a Supercritical Water Natural Circulation Loop (SCWNCL). A separate computer code NOLSTA has been developed which investigates stability characteristics of supercritical natural circulation loop using non-linear analysis. The conservation equations of mass, momentum and energy in transient form were solved numerically using finite volume method. The stable, unstable and neutrally stable points were identified by examining the amplitude of flow and temperature oscillations with time for a given set of operating conditions. The stability behavior of loop, predicted using non-linear analysis has been compared with that obtained from linear analysis. The results show that the stability maps obtained by the two methods agree qualitatively. The present paper describes the linear and nonlinear stability analysis models and the results obtained in detail.© 2009 ASME

Heat Transfer Studies on Lead-Bismuth Eutectic Flows in Circular Tubes

The use of accurate heat transfer model in liquid metal like Lead Bismuth Eutectic (LBE) flow is essential for the designing of the liquid metal cooled nuclear reactor systems. In the present study, the existing physical correlations for heat transfer in LBE flow through circular tube have been reviewed and assessed with the experimental results. In CFD analysis, PHOENICS-3.6 is used to carry out the evaluation of the various turbulence models in the tube geometry and to identify the difference between the numerical results and experimental ones in LBE flows. Based on the assessment of the existing correlations for heat transfer in LBE flow and the CFD results achieved, the best-suited correlation for turbulent Prandtl number is recommended in terms of Peclet number. This Prt can be incorporated in PHOENICS for LBE flow analysis.Copyright

A Comparative Study of Single-Phase, Two-Phase and Supercritical Natural Circulation in a Rectangular Loop

A one dimensional theoretical model has been used to analyze the steady state and stability performance of single-phase, two-phase and supercritical natural circulation in a uniform diameter rectangular loop. Parametric influences of diameter, inlet temperature and system pressure on the steady state and stability performance has been studied. In the single-phase liquid filled region, the flow rate is found to increase monotonically with power. On the other hand the flow rate in two-phase NCS is found to initially increase, reach a peak and then decrease with power. For the supercritical region also, the steady state behaviour is found to be similar to that of two-phase region. However, if the heater inlet temperature is beyond the pseudo critical value, then the performance is similar to single-phase loops. Also, the supercritical natural circulation flow rate decreases drastically during this condition. With increase in loop diameter, the flow rate is found to enhance for all the three regions of operation. Pressure has a significant influence on flow rate in two-phase region marginal effect in supercritical region and practically no effect in the single-phase region. With increase in loop diameter, operation in the single-phase and supercritical regions is found to destabilize whereas the two-phase loops are found to stabilize. Again, pressure has a significant influence on stability in the two-phase region.Copyright

Effect of Inlet Configuration on Moderator Velocity and Temperature Distribution Inside the Calandria of a Heavy Water Reactor

Computational study of the moderator flow in calandria vessel of a heavy water reactor is carried out for three different inlet nozzle configurations. For the computations, PHOENICS CFD code is used. The flow and temperature distribution for all the configurations are determined. The impact of moderator inlet jets on adjacent calandria tubes is studied. Based on these studies, it is found that the inlet nozzles can be designed in such a way that it can keep the impact velocity on calandria tubes within limit while keeping maximum moderator temperature well below its boiling limit.Copyright

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.