Stephen M. Bajorek

Experimental Investigation of Inverted Annular Film Boiling in a Rod Bundle during Reflood Transient

Abstract Heat transfer results for subcooled and saturated inverted annular film boiling (IAFB) obtained from a 7×7 rod bundle during transient reflood are presented in this paper. The test section consists of heater rods of 9.5-mm diameter and 12.6-mm pitch arranged in a square array. Flooding rates considered are 0.076 and 0.152 m/s, pressure varied from 138 to 414 kPa, and inlet subcooling up to 83 K. Evaluation of the data includes estimation of the local void fraction and Nusselt number during IAFB as well as in the inverted slug film boiling (ISFB) regime, which occurs when the inverted annular liquid column disintegrates. Experimental heat transfer results are compared with several film boiling models, and a new correlation for the Nusselt number is proposed for the IAFB and ISFB regimes. Predicted Nusselt numbers using the new correlation deviate from the experimental data by an average error of 15% and root-mean-square error of ∼30%.

Assessment of TRACE Code Using Rod Bundle Heat Transfer Mixture Level-Swell Tests

The TRAC/RELAP Advanced Computational Engine (TRACE) thermal-hydraulics code is currently under development by the United States Nuclear Regulatory Commission (NRC). TRACE is used for safety analyses of both conventional and advanced light water reactors. NRC assessed the prediction accuracy of the code by quantifying the axial void distribution in a rod bundle under low-pressure (0.16 to 0.44 Pa) and low-flow conditions (0.015 to 0.20 kg/s), using data obtained from the Rod Bundle Heat Transfer (RBHT) facility at Pennsylvania State University. NRC simulated 73 steady-state experiments (assessment cases) with variations in the total rod power, inlet subcooling, system pressure, and injection flow rate. Comparisons between TRACE calculations and RBHT data showed reasonable agreement. TRACE was found to over predict the bundle-exit void fraction by 13.3% with a linear goodness-of-fit (R2 ) of 0.87 and over predict the local void fraction by 10.1% with an R2 of 0.91. This paper discusses the models and correlations used in the TRACE calculation of mixture level swell, RBHT experimental results, modeling of the RBHT facility, and comparisons between data and code calculations.Copyright

Liquid Entrainment at Tee Junction With Upward Oriented Vertical Branch

Investigation on liquid entrainment in tee junctions with an upward oriented vertical branch was conducted. Existing correlations and test data were identified for evaluation and comparison. Theoretical analysis was carried out for the development of a new entrainment onset criterion and an entrainment rate model. Comparisons were performed with all available data with different test conditions, and the results indicated that the new correlations demonstrated a marked improvement.Copyright

Experimental Studies of Spacer Grid Thermal Hydraulics in the Dispersed Flow Film Boiling Regime

Abstract Spacer grids have been found to enhance downstream convective heat transfer and to strongly influence droplet size distributions through early spacer grid rewet and droplet breakup. Existing models for enhancement of heat transfer and droplet breakup, however, do not appear to accurately account for these interactions between the coolant and the spacer grid. Data from two series of rod bundle heat transfer tests, low injection rate forced reflood tests, and droplet injection tests are presented in this paper to describe the effects of the spacer grids during dispersed flow film boiling. Heat transfer downstream of the spacer grids is clearly enhanced by the presence of the droplets, while the downstream droplet size was found to depend on the condition of the spacer grid: dry or wetted. Results of this study demonstrate the need to adequately account for the separate modes of dry and wet spacer grid heat transfer enhancement in predicting the thermal-hydraulic behavior during reflood transients.

Rod Bundle Heat Transfer Thermal-Hydraulic Program

Abstract The U.S. Nuclear Regulatory Commission has been conducting thermal-hydraulic research using the Rod Bundle Heat Transfer (RBHT) facility at the Pennsylvania State University since 2001. The facility has been used for five individual test programs: forced reflood, steam cooling, mixture level swell, dispersed droplet injection, and oscillatory reflood test series. While rod bundle thermal hydraulics has been extensively studied in the past, the RBHT data have provided new insights into rod bundle phenomena especially on the effects of spacer grids. This paper provides a summary of the RBHT test program and discusses some of the major findings from this research with the emphasis on reflood thermal hydraulics and the effect of spacer grids. Of particular interest are data that enable model and correlation development. Recent efforts have focused on the evaluation of RBHT data and development of improved models and correlations suitable for systems thermal-hydraulic codes such as TRACE and RELAP. Because of detailed instrumentation on and about spacer grids, RBHT data have enabled improved models for convective heat transfer enhancement and droplet breakup. New correlations for the inverted annular and the inverted slug film boiling regimes have also been developed as an initial step toward an improved model for dispersed droplet film boiling.

Effects of Substrate Materials and Surface Conditions on the Minimum Film-Boiling Temperature

Abstract Film boiling is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident in a nuclear reactor. This study investigates the effects of liquid subcooling, surface oxidation, and surface materials on the minimum film-boiling temperature . Quenching experiments were performed using stainless steel and zirconium (Zr) test samples. The samples were heated to a temperature well above then plunged vertically in various degrees of liquid subcooling pools. A visualization study using a high-speed camera was conducted to capture the quenching behavior. Additionally, surface characterization analyses including X-ray diffraction, scanning electron microscopy, and energy dispersive X-ray spectroscopy were performed to quantify the surface conditions. Results indicate that liquid subcooling has a strong influence on . The visualization study shows a very thin vapor formation around the test sample for higher subcooling pools which explains the enhancement in the heat transfer. It is observed from the surface characterization analyses that the variations in the surface condition of the stainless steel and Zr causes the vapor bubbles to depart differently in the nucleate boiling regime. Furthermore, the effect of surface oxidation is clearly noticeable in the Zr test sample compared to the stainless steel test sample due to the oxidation kinematic of each substrate material. It is found that the substrate thermophysical properties have a significant impact on . Comparing the bare substrates shows that for the same degrees of liquid subcooling pool, the value of for the Zr sample is ∼30°C to 60°C higher compared to the stainless steel sample. Moreover, increasing the degrees of liquid subcooling contributes to a significant increase in that varies between ∼50°C and 70°C for both samples.