[en] Thermal hydraulic data-base is very useful in the design and analysis of the proposed Advanced Heavy Water Reactor which relies on natural circulation for normal core cooling. Compilation of the thermal hydraulic data-base is in progress. Artificial Neural Networks (ANNs), have been applied to analyse the consistency and accuracy of the data-base. The ANN predictions are more accurate and cover wider range of parameters compared to model based predictions

[en] Air-Water Loop (AWL) is an experimental facility aimed at the simulation of two-phase flow phenomenon relevant to steam drum of Advanced Heavy Water Reactor (AHWR). AWL consists of scaled down model of AHWR steam drum, operating with air-water mixture. Scaling has been performed such that the superficial velocities of both phases in the model are identical to that in the prototype. The required boundary flow, i.e. two phase air-water flow at drum inlet is generated using two-phase natural circulation loop formed by pipe channels connected between the drum and a water tank. Air is injected at the bottom of channels and is separated in the drum. Steady state experiments have been performed in AWL at various operating levels in drum and air injection flow rates. The experimental data has been validated using a numerical model developed based on momentum balance in channels. This paper describes the geometry of the experimental facility, details of two-phase natural circulation experiments carried out and validation of experimental data. (author)

[en] Full text: Many upcoming new generation reactors are natural circulation based reactor with multiple vertical fuel channels with on power refueling/defuelling. Natural circulation systems are simpler and safer than forced circulation systems. However, natural circulation systems are susceptible to instabilities. The parallel channel under natural circulation makes the system more complicated.To examine the behavior of parallel channel system, analysis was carried out for natural circulation flows 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. The flow behavior of parallel channel system for different rate of change of heater power is also investigated. The system under consideration consists of three parallel heated channels along with down-comer connected with common inlet and outlet headers. Flow behavior of the system was investigated for different cases of single and multiple heated channels in which all possible cases of single and twin heated channels were considered. Similarly, different cases were considered to study the effect of different rate of change of heater power on the behavior of the multiple heated channels system. The paper brings out the details of system considered, cases analyzed and results of analysis carried out on single phase parallel channel system using RELAP5/MOD3.2

[en] The Indian Pressurized Heavy Water Reactor (IPHWR) is a heavy water moderated and heavy water cooled nuclear reactor which employs short fuel bundles stacked one after another inside each of the horizontal fuel channels. The 540 MWe reactors, under operation in India, employ 37- rod bundles. Upcoming IPHWR designs of 700 MWe rating allow boiling of the coolant at the exit of channels. Therefore, there is a two-phase flow near at the exit of fuel channel, which causes the pressure drop to increase in the channel due to its two-phase conditions. In the absence of two-phase pressure drop data for 37-rod bundle, it is of prime interest and requirement to estimate the pressure drop across 37- rod fuel bundle under two-phase conditions. The experiments have been conducted in existing Boiling Water Loop (BWL) for different flows and qualities at a pressure of 70 bar across the simulated fuel channel of 700 MWe IPHWR. The data generated for pressure drop across fuel bundle have been analyzed to evaluate the two-phase friction multipliers. This paper deals with the experimental set-up, the experiments carried out and the analysis of the results in details. (author)

[en] Pressurized Heavy Water Reactor (PHWR) is a heavy water moderated and heavy water cooled nuclear reactor which employs short fuel bundles stacked one after another inside each of the horizontal channels. The 540 MWe reactors, under operation in India, employ 37- rod bundles. Presently, up rating of 540 MWe reactors to 700 MWe is taken up which allows boiling of the coolant, at the exit of channels. However, this provision affects the pressure drop of the coolant as it passes through different geometries, as pressure drop is a function of flow geometry besides other parameters. It is of prime interest and requirement to estimate the pressure drop across 37-rod fuel bundle under two-phase conditions. In the present study, experiments have been conducted in existing 3 MW Boiling Water Loop (BWL) at BARC for different flows and qualities at a pressure of 70 bar across the simulated fuel channel of 700 MWe PHWR. The results obtained for pressure drop across fuel bundle have been analyzed to evaluate the two-phase friction multipliers. (author)

[en] Supercritical pressure natural circulation experiments were carried out with CO₂ in a uniform diameter rectangular loop. Experimental data were generated on, steady state flow, heat transfer and stability under natural circulation conditions. The steady state flow rate data obtained were compared with the predictions of 1-D code NOLSTA which showed good agreement. The supercritical heat transfer coefficient data showed a peak around the pseudocritical point. The heat transfer coefficient data were compared with different correlations reported in the literature. Good agreement was obtained with the prediction of McAdams, Bishop, Jackson and Jackson Fewester correlations. Instability was observed in the loop in a narrow window around the pseudo critical region with low cooling water flow rate for the HHHC orientation. All other orientations of heater and cooler were found to be stable. The stability data were compared with the predictions of the nonlinear stability analysis code NOLSTA. The details of the experimental set-up, experiments carried out and the results of the analysis are presented in this report. (author)

[en] Earlier, 1/2 ″ uniform diameter Supercritical Pressure Natural Circulation Loop (SPNL) was set-up in hall-7, BARC for carrying out experiments related to supercritical fluids. The loop is a rectangular loop having two heaters and two coolers. Experiments were carried out with CO₂ under supercritical conditions for various pressures and different combinations of heater and cooler orientations. Since, the design conditions are more severe for supercritical water (SCW) experiments, the loop was modified for SCW by installing new test sections, pressurizer and power supply for operation with supercritical water. Experimental data were generated on steady state, heat transfer and stability under natural circulation conditions for the horizontal heater and horizontal cooler (HHHC) orientation with SCW up to a heater power of 8.5 kW. The flow rate data and instability data were compared with the predictions of in-house developed 1-D code NOLSTA, which showed reasonable agreement. The heat transfer coefficient data were also compared with the predictions of various correlations exhibit peak at bulk temperature lower than that obtained in the experiments. Most of these correlations predicted experimental data well in the pseudo-critical region. However, all correlations are matching well with experimental data beyond the pseudo-critical region. The details of the experimental facility, Experiments carried out and the results presented in this report. (author)

[en] Design of an Advanced Heavy Water Reactor (AHWR) is in progress in India to enable effective thorium utilization. The proposed system envisages a pressure tube type of heavy water moderated reactor with vertical channels using boiling light water as coolant. The reactor core consists of Th-U233 fuel driven by plutonium-uranium mixed oxide driver assemblies. As part of the core design exercise, preliminary thermal hydraulic analysis of the core has been carried out. Core flow distribution, among other parameters, is largely dependent upon hydraulic resistances at channel inlets. This, in turn, influences the steam quality at channel exit. Analysis has been carried out for different values of inlet resistances. One of the passive safety features proposed to be incorporated in the AHWR is to maintain the flow of coolant through the core by natural circulation. Two phase thermosyphon analysis has been carried out to study the effect of variation of loop height and core power on core flow. Details of the analyses carried out along with the results are discussed in the paper. (author). 6 refs, 5 figs, 2 tabs

[en] Pressure drop is an important parameter for design and analysis of many systems and components. For example, validated pressure drop correlations are required to determine the extent of orificing needed at core inlet, pumping power required, the riser height required in natural circulation boiling water reactors (BWRs), recirculation ratio in steam generators, etc. Some of the above applications require correlations for both single-phase and two-phase flows

[en] In the proposed Indian Advanced Heavy Water Reactor (AHWR) the coolant recirculation in the primary system is achieved by two-phase natural circulation. The two-phase steam-water mixture from the reactor core is separated in steam drum by gravity. Gravity separation of phases may lead to undesirable phenomena - carryover and carryunder. Carryover is the entrainment of liquid droplets in the vapor phase.Carryover needs to be minimized to avoid erosion corrosion of turbine blades. Carryunder is the entrainment of vapor bubbles with liquid flowing back to reactor core. Significant carryunder may in turn lead to reduced flow resulting in reduced CHF margin and stability in the coolant channel. An Air-Water Loop (AWL) has been designed to carry out the experiments relevant to AHWR steam drum. The design features and scaling philosophy is described in this report. (author)