Search

The LOCA (loss of coolant accident) is a kind of severe accident in the operation of PHWR (pressurized heavy water reactor) as well as other nuclear facilities, and possible cause of LOCA can be counted on the ballooning of pressure tube (PT) contacted to the outer calandria tube (CT) in the moderator system of CANDU-6 reactors. In the paper, we simulated the 150-kW experimental facility proposed by IAEA/ISCP, modeling the transient creeping behavior of pressurized tube heated with thermal radiation between the gaps of two concentric pipes. The outer boundary is simplified with a switched model that depends on the local temperature. With a multiphysical model supported by a commercial code, COMSOL multiphysics, the unsteady phenomena are simulated with models concerning various kinds of mechanics such as thermodynamics, nonlinear structural dynamics, and two-phase boiling heat transfer models. [ABSTRACT FROM AUTHOR]

Abstract: The results of two Small Break Loss of Coolant Accident (SBLOCA) experiments in RD-14M test facility and their predictions by CATHENA code, which is used to analyze postulated events in CANDU reactors, are compared in this paper. Two specific SBLOCA experiments selected for the CATHENA code predictions are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow-down. These blind simulations demonstrate that CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. The CATHENA predictions of fuel sheath temperatures for test B9006 are in much better agreement with the test measurements than those for test B9802, because CATHENA code could not capture the oscillatory behavior of channel flow and consequently sheath temperature when local fuel sheath surface was being intermittently dried and rewet under near stagnant flow and full power conditions of test B9802. Therefore, it is concluded that further works are required to appropriately predict the sheath temperature spike and its fluctuation during the transient of test B9802 where the critical heat flux and post-dryout heat transfer are important governing phenomena. [ABSTRACT FROM AUTHOR]

Abstract: To form a licensing basis for a new methodology for a fuel channel safety analysis code for CANDU-6 nuclear reactor, a CATHENA model for a post-blowdown fuel channel analysis has been developed, and tested for a high temperature thermal–chemical experiment CS28-1 [Lei, Q.M., 1993. Post-test analysis of the 28-element high-temperature thermal–chemical experiment CS28-1. In: 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal, PQ, 1993]. Pursuant to the objective of this investigation, the current study has focused on understanding the involved phenomena, their interrelations, and how to maintain a good accuracy of the temperature and H2 generation rate prediction without losing the important physics of the involved phenomena. The transient simulation results for the fuel element simulators (FESs) for the three fuel rings and the pressure tube were reasonably good as proven by the simulation results. This is thought to be due to success in reproducing the initial steady state condition well, and due to the detailed modeling features of CATHENA code for the coupled conduction, convection, and radiation heat transfer in a complex fuel bundle geometry. However, one problem still remained unresolved, i.e. the inability to accurately predict the pressure tube temperature at the initial steady state condition, and an adjustment of the CO2 gap conductivity necessary to match the measured pressure tube temperatures. However, this raises a question as to how the transient FES and the pressure tube temperature can be predicted so well in spite of an insufficient justification for using the “non-participating medium assumption” for the CO2 gas gap. Through this study and the affiliated previous study for the steady state, it was found that the radiation heat transfer model of CATHENA among the FESs of the three rings and the pressure tube as well as the exothermic metal–water reaction model based on the Urbanic–Heidrick correlation are reasonably accurate and sound. Also it was found that an accurate prediction of the initial condition of the experiment is very important for an accurate prediction of the whole transient as it serves as the starting point for the transient. [Copyright &y& Elsevier]

Abstract: To form a licensing basis for the new methodology of the fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal–chemical experiment CS28-1. As the major concerns of the post-blowdown fuel channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H2 at how high a fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20–25°C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using “transparent” assumption for the CO2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA code’s inherent modeling feature is raised from this study. Based on a comparison of the prediction of the current CATHENA model and the experiment data, the steady state model is deemed to be adequate as a starting point for the following high temperature thermal–chemical experiment of a metal–water reaction. [Copyright &y& Elsevier]

Abstract: A computational fluid dynamics (CFD) model of a post-blowdown fuel channel analysis for aged CANDU reactors with crept pressure tube has been developed, and validated against a high temperature thermal–chemical experiment: CS28-2. The CS28-2 experiment is one of three series of experiments to simulate the thermal–chemical behavior of a 28-element fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA (large break loss of coolant accident) of CANDU reactors. Pursuant to the objective of this study, the current study has focused on understanding the involved phenomena such as the thermal radiation and convection heat transfer, and the high temperature zirconium-steam reaction in a multi-ring geometry. Therefore, a zirconium-steam oxidation model based on a parabolic rate law was implemented into the CFX-10 code, which is a commercial CFD code offered from ANSYS Inc., and other heat transfer mechanisms in the 28-element fuel channel were modeled by the original CFX-10 heat transfer packages. To assess the capability of the CFX-10 code to model the thermal–chemical behavior of the 28-element fuel channel, the measured temperatures of the fuel element simulators (FES) of three fuel rings in the test bundle and the pressure tube, and the hydrogen production in the CS28-2 experiment were compared with the CFX-10 predictions. In spite of some discrepancy between the measurement data and CFX results, it was found that the CFX-10 prediction based on the Urbanic–Heidrick correlation of the zirconium-steam reaction as well as the Discrete Transfer Model for a radiation heat transfer among the FES of three rings and the pressure tube are quite accurate and sound even for the offset fuel cluster of an aged fuel channel. [Copyright &y& Elsevier]

Abstract: In this study a CFD simulation of a horizontal heater rods test (CS28-2) has been performed with the CFX-10 code and it has been compared with an experiment in order to develop a post-blowdown modeling of a CANDU fuel channel during a large break loss of coolant accident (LBLOCA). The CS28-2 experiment is one of three series of experiments to simulate the fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA of CANDU reactors. Since radiation heat transfer is dominant in the CS28-2 experiment, a benchmark problem of radiation heat transfer for the same geometry as the CS28-2 is carried out first to assess the radiation model by CFX-10, to confirm whether CFX-10 can be applied to a complex fuel geometry or not. And, the CFD study of the CS28-2 has been performed with the aim to simulate the steady-state condition of the CS28-2. These steady-state calculation results will be used for the initial conditions of the transient simulation. [Copyright &y& Elsevier]

Three-dimensional moderator flow in the calandria tank of CANDU-6 pressurized heavy water reactor (PHWR) is computed with Open Field Operation and Manipulation (OpenFOAM), an open-source computational fluid dynamics (CFD) code. In this study, numerical analysis is performed on the real geometry model including 380 fuel rods in the calandria tank with the heat-source distribution to remove uncertainty of the previous analysis models simplified by the porous media approach. Realizable k-ε turbulence model is applied, and the buoyancy due to temperature variation is considered by Boussinesq approximation for the incompressible single-phase Navier-Stokes equations. The calculation results show that the flow is highly unsteady in the moderator. The computational flow visualization shows a circulation of flow driven by buoyancy and asymmetric oscillation at the pseudo-steady state. There is no region where the local temperature rises continuously due to slow circulating flow and its convection heat transfer. [ABSTRACT FROM AUTHOR]

In the present study, the natural convection heat transfer of liquid gallium is investigated in a rectangular loop which consists of an indirect heating block test section for heat transfer rates, 1/2” tubes and a condenser as well as an orifice for measuring flow rate. The mass flow rate for natural convection and average Nusselt number of liquid gallium were measured within the heat flux range of 6.17×103–5.07×104 W/m2. The measured heat transfer rates were correlated as follows: corresponding to laminar natural convection regime. Also, the flow rates for the natural convection of liquid gallium depending on power level were also compared by using a CFD code and MARS-LMR (Multi-dimensional Analysis of Reactor Safety) code. The MARS code was modified into so-called MARS-Ga applicable for gallium-cooled systems in terms of physical properties. It was found that the predictions were in good agreements of the trend with the experimental results. [ABSTRACT FROM AUTHOR]