Simulation of Critical Heat Flux Experiments in Neptune_CFD Code (original) (raw)
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CFD SIMULATION OF CRITICAL HEAT FLUX IN A TUBE
This paper presents numerical simulations of the boiling flow in a tube with a Departure from Nucleate Boiling type of critical heat flux (CHF). Standard tables of CHF produced by the Russian Academy of Sciences were used as a data set. The simulations were performed with the multiphase code NEPTUNE_CFD V1.0.7. A simple criterion based on the void fraction at the wall was used for the CHF prediction. Four data series were selected from the tables. In every series, one of the following parameters was variable: the local equilibrium quality, the mass flux, pressure and the tube diameter. The remaining three parameters were fixed. In every data point, a numerical simulation was performed so as to find out the interval of the wall heat fluxes at which the boiling crisis occurs. NEPTUNE was able to quite accurately predict CHF in cases with high mass fluxes and high pressures. On the other hand, in one low-mass-flux case, the CHF in the calculation occurred at a wall heat flux as low as ...
Simulation of Boiling Flow Experiments Close to CHF with the Neptune_CFD Code
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A three-dimensional two-fluid code Neptune_CFD has been validated against the Arizona State University (ASU) and DEBORA boiling flow experiments. Two-phase flow processes in the subcooled flow boiling regime have been studied on ASU experiments. Within this scope a new wall function has been implemented in the Neptune_CFD code aiming to improve the prediction of flow parameters in the near-wall region. The capability of the code to predict the boiling flow regime close to critical heat flux (CHF) conditions has been verified on selected DEBORA experiments. To predict the onset of CHF regime, a simplified model based on the near-wall values of gas volume fraction was used. The results have shown that the code is able to predict the wall temperature increase and the sharp void fraction peak near the heated wall, which are characteristic phenomena for CHF conditions.
Analytical and CFD Investigation of Ex-Core Cooling of the Nuclear Fuel Rod Bundle in a Water Pool
12th International Conference on Nuclear Engineering, Volume 2, 2004
A long 778 km high pressure, submarine pipeline supplying natural gas will serve multiple combined cycle gas turbine peaking units on the Pacific Rim. The gas supply flow into the pipeline is constant, but the power plant will primarily operate during daylight hours or in certain situations, operate in a two-shift mode. So the pipeline essentially serves as an increasing pressure, gas storage vessel during the night, and pressure falls o ff during the day as the gas is fired. Hence, the gas letdown receiving station presented many critical design challenges. Among these were the need for constant plant service pressure control, especially during individual power generator unit startup, shutdown, and upset conditions. In addition, there was a very severe noise attenuation requirement and high gas flow rangeability was required. This complex, integrated gas pressure, control valve letdown system and its operation is described in detail.
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Critical heat flux (CHF) tests have been performed for a vertically upward R134a flow in a 5 · 5 rod bundle in the following parameter ranges: outlet pressure 1.5-3.0 MPa, mass flux 50-2500 kg m À2 s À1 , inlet quality À0.08 to À0.75 and critical quality À0.17 to 1.58. Parametric trends of the CHF data agree well with the general understanding. Water equivalent CHF data is generated using a set of well-established modeling parameters, then compared with (a) the EPRI/Colombia University's water CHF database for similar geometry and experimental conditions and (b) commonly used CHF prediction methods. The water equivalent CHF data generated from the present tests shows a good agreement with the actual water CHF data for both Katto's and Ahmad's modeling methods when considering the differences in the detailed geometric and flow conditions. They are also predicted well by the 1995 CHF look-up table. These results indicate that R134a can be a good modeling fluid for the CHF of water in rod bundles at least in the investigated parameter ranges.
CFD analysis of flow boiling in the ITER first wall
Fusion Engineering and Design, 2012
This paper compares two Computational Fluid Dynamic (CFD) approaches for the analysis of flow boiling inside the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER): (1) the Rohsenow model for nucleate boiling, seamlessly switching to the Volume of Fluid (VOF) approach for film boiling, as available in the commercial CFD code STAR-CCM+, (2) the Bergles-Rohsenow (BR) model, for which we developed a User Defined Function (UDF), implemented in the commercial code FLUENT. The physics of both models is described, and the results with different inlet conditions and heating levels are compared with experimental results obtained at the Efremov Institute, Russia. The performance of both models is compared in terms of accuracy and computational cost.
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Critical heat flux (CHF) has been studied for almost a century and yet there is no indisputable consensus reached on governing physical phenomena behind, not to mention, on modelling agreement of different correlations. When we are compelled to run our system at the safe distance from the CHF, and we can use all the accumulated knowledge so far, we will quite possibly cling to look-up tables delivered with that particular system. If this is not the case, than we will certainly stick to the system-specific correlation, which cannot be applied with confidence elsewhere. In the last two decades there were significant advancements applied both in numerical simulation capabilities and in unintrusive measuring techniques, which shed light on anticipated advancements in modelling the phenomenon. However, there are few reliable experimental measurements of instantaneous velocity and temperature fields in the wall boundary layer, and they are nil where local heat transfer coefficients are acquired. Therefore, at Reactor Engineering Division of Jožef Stefan Institute, a unique test section for local critical heat flux studies is under construction. The selected geometry and the test conditions will resemble light water reactorlike accident conditions. Moreover, to understand the phenomenon better, the design of the test section enables local measurements of heat transfer coefficients, and allows for control over the diabatic wall temperature. Measurements of single-phase convective heat transfer, conjugate heat transfer, flow boiling, convective condensation, and condensation-induced liquid hammer were all part of the test section's design basis. In this context, the design and construction of the device is herein presented in considerable detail.
Review of pool boiling critical heat flux (CHF) and heater rod design for CHF experiments in TREAT
Progress in Nuclear Energy, 2020
Preventing the occurrence of a departure from nucleate boiling (DNB) event is an important aspect of nuclear safety in pressurized water reactors (PWRs). This phenomenon is partially governed by the cladding-to-coolant heat transfer under transient irradiation conditions, such as during a reactivity-initiated accident (RIA). Currently, there are large uncertainties about cladding-to-coolant heat transfer under these rapid transient conditions. This effort aims to elucidate the mechanisms of CHF under in-pile fast transient irradiation conditions using the Transient Reactor Test (TREAT) facility. These experiments will be carried out under pool boiling conditions, within experimental capsules that will be inserted into the TREAT reactor. A heater rodlet made from stainless steel 304 with tailored natural boron content will be usedin these experiment capsules to induce CHF when submitted to a TREAT power pulse. We will investigate the impacts of the presence of an oxide layer, radiation-induced surface activation (RISA), heat transfer time constant, and rapid surface heating effects on the CHF phenomenon. We review the parameters with significant effects governing the predictions of pool boiling CHF. Only CHF influencing parameters that are highly important, among them coolant subcooling and pressure, as well as oxide layer thickness, RISA, and rapid heating effects were included in this literature assessment. Preliminary neutronics and thermal hydraulic results from the design of the experimental apparatus are also presented in this paper. To aid in the modeling approach, the energy deposition and occurrence of CHF were identified as the most crucial key Figures of Merit (FoMs). In the heater rod design, boron concentrations between 0.1 and 2.09 wt% were considered. Further, a self-shielding study was performed to determine whether an instrumented borated tube could be used in place of a solid borated rod. This study determined that the inner region of the rod can be excluded or instrumented without heat generation penalties. Lastly, a thermal-hydraulics sensitivity study determined the lowest limiting boron concentration needed to induce CHF in water with different degrees of subcooling. Additionally, the value of CHF is known to increase during a rapid transient. Therefore, a CHF multiplier sensitivity study determined what multipliers would inhibit CHF for varying degrees of subcooling of two chosen power coupling factors (PCFs). The current borated tube rodlet geometry configuration achieved a maximum CHF multiplier value of 7.8 using a 1400 MJ power pulse in TREAT. Although this was the power pulse with the greatest energy deposition considered for this study, the TREAT facility is capable of pulses up to ~2500 MJ. This provides a significant margin in energy capacity that was not included within the scope of the calculations carried out to date. The application of the heater rod design was successfully demonstrated in initial experiments in December 2019. The results of these experiments will be explored in future publications.
NEPTUNE_CFD Analysis of Flow Field in Rectangular Boiling Channel
The Journal of Computational Multiphase Flows, 2012
The boiling flow experiments in a rectangular vertical channel, performed at the Texas A&M University, were used to validate the prediction capabilities of the NEPTUNE_CFD code. The refrigerant HFE-301 was used as a working fluid. Liquid velocity and turbulent kinetic energy profiles close to the heated wall were compared. NEPTUNE_CFD simulations successfully predict the main experimental tendencies associated with the heat flux and Reynolds number variation. Nevertheless, at lower Reynolds numbers, a somewhat higher disagreement in velocity and turbulent kinetic energy can be observed in the boiling region. The size of bubble diameter at departure from the heated wall is assessed by the means of non-dimensional analysis and compared with model predictions.
Effect of Mass Flow Rate on Critical Heat Flux in Vertical Tube
Al-Nahrain Journal for Engineering Sciences, 2012
The critical heat flux (CHF) in forced convective up flow has been investigated in a uniformly heated vertical steel tube of 18 mm internal diameter and 3.66 m length, with water as the working fluid. The CHF was determined by the sudden rise in the wall temperature of the electrically heated tube. Experiments were performed at nominal pressure 50 bar over three mass flux values 0.11kg/s, 0.055kg/s and 0.027 kg/s, and heat flux with up to 400 kW maximum power. The CHF results in the present experimental ranges rise linearly as the mass flux increases. Finally, a comparison of the experimental data with available fluid correlations has been performed. The present results of CHF are found to be an acceptable agreement with those predicted by using (Bowring Correlation) and are slightly higher when compared with those predicted by Katto and Ohne Correlation . Introduction The critical heat flux (CHF) condition is characterized by a sharp reduction of the local heat transfer coefficient...
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