Evaluations of the Coolability Through the Inherent In-Vessel Gap Cooling in the LAVA Experiments

2002 ◽  
Author(s):  
K. H. Kang ◽  
R. J. Park ◽  
J. T. Kim ◽  
S. B. Kim ◽  
H. D. Kim
Keyword(s):  
Current Flow ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Gap Formation ◽  
Lower Head ◽  
Melt Composition ◽  
Counter Current ◽  
Vessel Bottom ◽  
Cooling Model

The analysis of the LAVA (Lower-plenum Arrested Vessel Attack) experimental results focused on gap formation and in-vessel gap cooling characteristics have been performed. In the LAVA experiment, Al2O3/Fe thermite melt (or Al2O3 only) was used as a corium simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation in case there was an internal pressure load across the vessel. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were mainly determined by the possibilities of the water ingression into the gap. The possibility of heat removal through the gap in the LAVA experiments was confirmed from that the vessel cooled down with the conduction heat flux through the vessel by 70 to 470 kW/m2. Also the quantitative evaluations of the in-vessel coolability using gap cooling model based on counter current flow limits (CCFL) have been performed for the LAVA experiments in parallel. It could be inferred from the analysis for the LAVA experiments that the vessel could effectively cooldown via heat removal through the gap cooling even if 2mm thick gap should form between the interface of the melt and the vessel in the 30 kg of Al2O3 melt tests. In the case of large melt mass of 70 kg of Al2O3 melt, however, the infinite possibility of heat removal through a small size gap such as 1 to 2 mm thick couldn’t be guaranteed due to the difficulties of water ingression through the gap into the lower head vessel bottom induced by the CCFL. Synthesized the experimental results and the analytical evaluations using the CCFL model, it could be found that the coolability through gap cooling was affected mainly by the melt composition and mass and also the gap thickness.

Experimental Study on the Improved In-Vessel Corium Retention Concepts for the Severe Accident Management

2002 ◽  
Author(s):  
K. H. Kang ◽  
R. J. Park ◽  
K. M. Koo ◽  
S. B. Kim ◽  
H. D. Kim
Keyword(s):  
Heat Removal ◽  
Elevated Pressure ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Severe Accident ◽  
Experimental Facility ◽  
Dual Strategy ◽  
Accident Management ◽  
Lower Head ◽  
Inner Diameter

Feasibility experiments were performed for the assessment of improved In-Vessel Corium Retention (IVR) concepts using an internal engineered gap device and also a dual strategy of In/Ex-vessel cooling using the LAVA experimental facility. The internal engineered gap device made of carbon steel was installed inside the LAVA lower head vessel and it made a uniform gap with the vessel by 10 mm. In/Ex-vessel cooling in the dual strategy experiment was performed installing an external guide vessel outside the LAVA lower head vessel at a uniform gap of 25 mm. The LAVA lower head vessel was a hemispherical test vessel simulated with a 1/8 linear scale mock-up of the reactor vessel lower plenum with an inner diameter of 500 mm and thickness of 25 mm. In both of the tests, Al2O3 melt was delivered into about 50K subcooled water inside the lower head vessel under the elevated pressure. Temperatures of the internal engineered gap device and the lower head vessel were measured by K-type thermocouples embedded radially in the 3mm depth of the lower head vessel outer surface and in the 4mm depth of the internal engineered gap device, respectively. In the dual strategy experiment, the Ex-vessel cooling featured pool boiling in the gap between the lower head vessel and the external guide vessel. It could be found from the experimental results that the internal engineered gap device was intact and so the vessel experienced little thermal and mechanical attacks in the internal engineered gap device experiment. And also the vessel was effectively cooled via mutual boiling heat removal in- and ex-vessel in the dual strategy experiment. Compared with the previous LAVA experimental results performed for the investigation of the inherent in-vessel gap cooling, it could be confirmed that the Ex-vessel cooling measure was dominant over the In-vessel cooling measure in this study. It is concluded that the improved cooling measures using a internal engineered gap device and a dual strategy promote the cooling characteristics of the lower head vessel and so enhance the integrity of the vessel in the end.

Experiments on Air-Water Countercurrent Flow in a Rectangular Duct Simulating PWR Hot Leg

2008 ◽  
Author(s):  
Noritoshi Minami ◽  
Daisuke Nishiwaki ◽  
Hironobu Kataoka ◽  
Akio Tomiyama ◽  
Shigeo Hosokawa ◽  
...  
Keyword(s):  
Current Flow ◽  
Heat Removal ◽  
Flow Patterns ◽  
Countercurrent Flow ◽  
Rectangular Duct ◽  
Horizontal Section ◽  
Pressurized Water ◽  
Mist Flow ◽  
Counter Current Flow ◽  
Counter Current

In the case of loss of the residual heat removal system under mid-loop operation during shutdown of the pressurized water reactor (PWR) plant, steam generated in a reactor core and condensed water in a steam generator (SG) form a countercurrent flow in a hot leg. In this study, in order to improve a counter-current flow model of a transient analysis code, experiments were conducted using a scale-down model of the PWR hot leg, and flow patterns and counter-current flow limitation (CCFL) characteristics were measured. A rectangular duct, whose height is about 1/5th of the hot leg diameter, was used to simulate the hot leg, and air and water at atmospheric pressure and room temperature were used for gas and liquid phases. In the horizontal section, as air flow rate QG increases, the flow pattern transits from a stratified flow to wavy flow, and then wavy to wavy-mist flow. When the latter transition takes place, water flow from the horizontal duct to the lower tank is to be restricted. Flow patterns in the elbow section are the same as those in the horizontal section. Wavy flow is not formed in the inclined section, where the transition to wavy-mist flow occurs due to the inflow of wavy-mist flow generated in the horizontal section. Flow patterns in the elbow and inclined section are strongly affected by those in the horizontal section. CCFL characteristics are well correlated with the Wallis-type correlation, and the onset of CCFL well corresponds to the transition from wavy flow to wavy-mist flow.

Comparison of CCFL Experiments Performed in Two Different Models of the Hot Leg of a PWR With Rectangular Cross-Section

2010 ◽  
Author(s):  
Christophe Valle´e ◽  
Tobias Seidel ◽  
Dirk Lucas ◽  
Akio Tomiyama ◽  
Michio Murase
Keyword(s):  
Cross Section ◽  
Flow Rate ◽  
Current Flow ◽  
Characteristic Length ◽  
Rectangular Cross Section ◽  
Experimental Results ◽  
Channel Height ◽  
Flow Limitation ◽  
Counter Current Flow ◽  
Counter Current

In order to investigate the two-phase flow behaviour during counter-current flow limitation in the hot leg of a pressurised water reactor, two test models were built: one at the Kobe University and the other at the TOPFLOW test facility of Forschungszentrum Dresden-Rossendorf (FZD). Both test facilities are devoted to optical measurement techniques, therefore, a flat hot leg test section design was chosen. Counter-current flow limitation (CCFL) experiments were performed, simulating the reflux condenser cooling mode appearing in some accident scenarios. The fluids used were air and water, both at room temperature. The pressure conditions were varied from atmospheric at Kobe to 3.0 bar absolute at TOPFLOW. According to the presented review of the literature, very few data is available on flooding in channels with rectangular cross-section, and no experiments were performed in the past in such rectangular models of a hot leg. Usually, the macroscopic effects of CCFL are represented in a flooding diagram, where the gas flow rate is plotted versus the discharge water flow rate. Commonly, the non-dimensional superficial velocity (also known as the Wallis parameter) is used to plot the flooding diagram. However, the classical definition of the Wallis parameter contains the pipe diameter as characteristic length, which was originally defined by Wallis (1969) for counter-current flow limitation in vertical pipes and not in near horizontal channels with rectangular cross-section. In order to be able to perform comparisons with pipe experiments and to extrapolate to the power plant scale, the appropriate characteristic length should be determined. Because the experimental projects on this subject at the Kobe University and at FZD were launched independently, a detailed comparison of both test facilities is presented. With respect to the CCFL behaviour, it is shown that the essential parts of the two hot leg test sections are very similar. This geometrical analogy allows to perform meaningful comparisons. However, clear differences in the dimensions of the cross-section (H × W = 150 × 10 mm2 in Kobe, 250 × 50 mm2 at FZD) make it possible to point out the right characteristic length for hot leg models with rectangular cross-sections. The hydraulic diameter, the channel height and the Laplace critical wavelength (leading to the Kutateladze number) were tested. The experimental results obtained in the two test facilities clearly show that the channel height is the suited characteristic length. Finally, the experimental results are compared with similar experiments and empirical correlations for pipes available in the literature. In spite of the scatter of the data and of the different correlations, it was noticed that flooding is reached at slightly lower gas fluxes in the hot leg models with rectangular cross-section compared to pipes.

Modeling Single-Phase Counter-Current Natural Convection Heat Removal in Horizontal Heated Channel Connected to Vertical Piping

2005 ◽  
Author(s):  
P. Gulshani ◽  
H. M. Huynh
Keyword(s):  
Water Flow ◽  
Single Phase ◽  
Current Flow ◽  
Heat Removal ◽  
Horizontal Channel ◽  
Intermittent Flow ◽  
Two Phase ◽  
Subcooled Water ◽  
Counter Current Flow ◽  
Counter Current

This paper develops a simple mathematical model to examine the heat transfer phenomena in a single-phase counter-current subcooled water flow in a volumetrically heated horizontal channel connected to an unheated vertical pipe at each end as shown in Figure 1. In Figure 1, the heated horizontal channel and the vertical pipes connected to it are initially filled with subcooled water up to a certain height in the vertical pipes. The vertical pipes can have horizontal runs. The piping arrangement in the model with horizontal fuel (i.e., heated) channels and vertical feeder pipes is relevant to a reactor such as the Canadian Deuterium Uranium (CANDU) reactor. The single-phase water flow condition considered in the model is relevant to CANDU in a shutdown, maintenance state where the main heat-transport-circuit pumps are shutoff and the shutdown-cooling pumps are or become unavailable. Under such postulated loss-of shutdown-cooling pump scenario, it is desirable to know whether the fuel fission-product decay heat can be adequately removed by single-phase subcooled water natural-circulation flow before the water in the fuel channels begins to boil. Boiling and the resulting two-phase conditions, condensation and changes in the buoyancy forces induce intermittent flow in the channel causing intermittent limited fuel heatup Ref [1–3]. Unlike counter-current flow of gas and liquid, counter-current flow of liquids, particularly the same miscible unequal-temperature liquids and in the geometry considered in this paper has not been studied either theoretically or experimentally to the authors’ knowledge.

Prediction of Severe Accident Counter Current Natural Circulation Flows in the Hot Leg of a Pressurized Water Reactor

2006 ◽  
Author(s):  
Christopher F. Boyd
Keyword(s):  
Steam Generator ◽  
Discharge Coefficient ◽  
Current Flow ◽  
Natural Circulation ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Flow Rates ◽  
Surge Line ◽  
Counter Current Flow ◽  
Counter Current

During certain phases of a severe accident in a pressurized water reactor (PWR), the core becomes uncovered and steam carries heat to the steam generators through natural circulation. For PWR’s with U-tube steam generators and loop seals filled with water, a counter current flow pattern is established in the hot leg. This flow pattern has been experimentally observed and has been predicted using computational fluid dynamics (CFD). Predictions of severe accident behavior are routinely carried out using severe accident system analysis codes such as SCDAP/RELAP5 or MELCOR. These codes, however, were not developed for predicting the three-dimensional natural circulation flow patterns during this phase of a severe accident. CFD, along with a set of experiments at 1/7th scale, have been historically used to establish the flow rates and mixing for the system analysis tools. One important aspect of these predictions is the counter current flow rate in the nearly 30 inch diameter hot leg between the reactor vessel and steam generator. This flow rate is strongly related to the amount of energy that can be transported away from the reactor core. This energy transfer plays a significant role in the prediction of core failures as well as potential failures in other reactor coolant system piping. CFD is used to determine the counter current flow rate during a severe accident. Specific sensitivities are completed for parameters such as surge line flow rates, hydrogen content, as well as vessel and steam generator temperatures. The predictions are carried out for the reactor vessel upper plenum, hot leg, a portion of the surge line, and a steam generator blocked off at the outlet plenum. All predictions utilize the FLEUNT V6 CFD code. The volumetric flow in the hot leg is assumed to be proportional to the square root of the product of normalized density difference, gravity, and hydraulic diameter to the 5th power. CFD is used to determine the proportionality constant in the range from 0.11 to 0.13 and termed a discharge coefficient. The value is relatively unchanged for typical surge line flow rates as well as the hydrogen content in the flow. Over a significant range of expected temperature differences for the steam generator and reactor vessel upper plenum, the discharge coefficient also remained consistent. The discharge coefficient is a suitable model for determining the hot leg counter current flow rates during this type of severe accident.

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