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.

CFD Analysis of Single Phase Counter-Current Buoyancy Driven Flows and its Applications to Passive Reactor Design

2014 ◽  
Author(s):  
Frederic Sebilleau ◽  
Anuj K. Kansal ◽  
Raad I. Issa ◽  
Simon P. Walker
Keyword(s):  
Current Flow ◽  
Natural Circulation ◽  
Three Dimensional ◽  
Local Heat ◽  
Complex Nature ◽  
Countercurrent Flow ◽  
Cold Fluid ◽  
Buoyancy Driven Flows ◽  
Counter Current Flow ◽  
Counter Current

Increasingly, nuclear plants rely on natural circulation, for both fault conditions and / or normal power removal. Prediction of such buoyancy-driven flows is needed. However, their complex nature leads to 3D effects in ‘wide’ geometries, making prediction impossible with system codes. Even in slender “pipe-like” geometries countercurrent flow of hot and cold fluid makes a one-dimensional simulation totally misleading. However, simply moving to a three-dimensional CFD treatment is not sufficient. The strong anisotropy of the turbulence and the coexistence of various flow regimes make the choice of an appropriate turbulence model difficult. Countercurrent flow in a pipe might occur when the “natural” buoyant flow was of hot fluid up the pipe, but a feature such as a local heat-sink (an un-insulated valve in the pipe, perhaps) acts as a source of cold fluid, which attempts to flow down the pipe as a counter-current flow. On a different scale, counter current flow such as this would occur for example inside the secondary containment. This countercurrent flow problem captures the complexities of most buoyant flows, and this provides a challenging model problem. In this paper, we describe the design and preliminary analysis of an experimental rig being built to study this. Initial CFD and experimental results are presented.

Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant

2009 ◽  
Author(s):  
Osamu Kawabata ◽  
Masao Ogino
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Reactor Core ◽  
Creep Damage ◽  
Reactor Vessel ◽  
Reactor System ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Hot Gas ◽  
Core Meltdown

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.

Air/Water Counter-Current Flow Experiments in a Model of the Hot Leg of a Pressurized Water Reactor

2009 ◽  
Vol 131 (2) ◽  
Author(s):  
Christophe Vallée ◽  
Deendarlianto ◽  
Matthias Beyer ◽  
Dirk Lucas ◽  
Helmar Carl
Keyword(s):  
Steam Generator ◽  
Pressure Vessel ◽  
Test Section ◽  
Current Flow ◽  
Flow Limitation ◽  
Two Phase ◽  
Pressurized Water ◽  
Counter Current Flow ◽  
Counter Current ◽  
Inlet Chamber

Different scenarios of small break loss of coolant accident for pressurized water reactors (PWRs) lead to the reflux-condenser mode in which steam enters the hot leg from the reactor pressure vessel (RPV) and condenses in the steam generator. A limitation of the condensate backflow toward the RPV by the steam flowing in counter current could affect the core cooling and must be prevented. The simulation of counter-current flow limitation conditions, which is dominated by 3D effects, requires the use of a computational fluid dynamics (CFD) approach. These numerical methods are not yet mature, so dedicated experimental data are needed for validation purposes. In order to investigate the two-phase flow behavior in a complex reactor-typical geometry and to supply suitable data for CFD code validation, the “hot leg model” was built at Forschungszentrum Dresden-Rossendorf (FZD). This setup is devoted to optical measurement techniques, and therefore, a flat test-section design was chosen with a width of 50 mm. The test section outlines represent the hot leg of a German Konvoi PWR at a scale of 1:3 (i.e., 250 mm channel height). The test section is mounted between two separators, one simulating the RPV and the other is connected to the steam generator inlet chamber. The hot leg model is operated under pressure equilibrium in the pressure vessel of the TOPFLOW facility of FZD. The air/water experiments presented in this article focus on the flow structure observed in the region of the riser and of the steam generator inlet chamber at room temperature and pressures up to 3 bar. The performed high-speed observations show the evolution of the stratified interface and the distribution of the two-phase mixture (droplets and bubbles). The counter-current flow limitation was quantified using the variation in the water levels measured in the separators. A confrontation with the images indicates that the initiation of flooding coincides with the reversal of the flow in the horizontal part of the hot leg. Afterward, bigger waves are generated, which develop to slugs. Furthermore, the flooding points obtained from the experiments were compared with empirical correlations available in literature. A good overall agreement was obtained, while the zero penetration was found at lower values of the gaseous Wallis parameter compared with previous work. This deviation can be attributed to the rectangular cross section of the hot leg model.

Modeling Improvements for System Code Evaluation of Inlet Plenum Mixing Under Severe Accident Conditions Using CFD Predictions

2010 ◽  
Author(s):  
Christopher Boyd ◽  
Kenneth Armstrong
Keyword(s):  
Natural Circulation ◽  
Nuclear Regulatory Commission ◽  
Mixing Model ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Dimensional System ◽  
Pressurized Water ◽  
Surge Line ◽  
Accident Conditions ◽  
Regulatory Commission

An updated mixing model is developed for application to system codes used for predicting severe accident-induced failures of steam generator (SG) U-tubes in a pressurized-water reactor. Computational fluid dynamics is used to predict the natural circulation flows between a simplified reactor vessel and the primary side of an SG during a hypothesized severe accident scenario. The results from this analysis are used to extend earlier experimental results and predictions. These new predictions benefit from the inclusion of the entire natural circulation loop between the reactor vessel upper plenum and the SG. Tube leakage and mass flow into the pressurizer surge line also are considered. The predictions are utilized as a numerical experiment to improve the basis for simplified models applied in one-dimensional system codes that are used during the prediction of severe accident natural circulation flows. An updated inlet plenum mixing model is proposed that accounts for mixing in the hot leg as well as the inlet plenum region. The new model is consistent with the predicted behavior and can account for flow into a side-mounted pressurizer surge line if present. Sensitivity studies demonstrate the applicability of the approach over a range of conditions. The predictions are most sensitive to changes in the SG secondary side temperatures or heat-transfer rates at the SG tubes. Grid independence is demonstrated through comparisons with previous models and by increasing the number of cells in the model. This work supports the U.S. Nuclear Regulatory Commission (NRC) studies of SG tube integrity under severe accident conditions.

scholarly journals Air-Water Tests on Counter-current Flow Limitation at Lower End of Vertical Pipes Simulating Lower Part of Steam Generator U-tube

2014 ◽  
Vol 28 (1) ◽  
pp. 62-70 ◽  
Author(s):  
Takayoshi KUSUNOKI ◽  
Taiga DOI ◽  
Yuki FUJII ◽  
Takashi TSUJI ◽  
Michio MURASE ◽  
...  
Keyword(s):  
Steam Generator ◽  
Current Flow ◽  
Flow Limitation ◽  
Counter Current Flow ◽  
Counter Current

Air/Water Counter-Current Flow Experiments in a Model of the Hot Leg of a Pressurised Water Reactor

2008 ◽  
Author(s):  
Christophe Valle´e ◽  
Deendarlianto ◽  
Matthias Beyer ◽  
Dirk Lucas ◽  
Helmar Carl
Keyword(s):  
Steam Generator ◽  
Pressure Vessel ◽  
Test Section ◽  
Current Flow ◽  
Two Phase ◽  
Counter Current Flow ◽  
Counter Current ◽  
Flow Experiments ◽  
Inlet Chamber ◽  
Reactor Pressure

Different scenarios of small break Loss of Coolant Accident (LOCA) for pressurised water reactors (PWR) lead to the reflux-condenser mode in which steam enters the hot leg from the reactor pressure vessel (RPV) and condenses in the steam generator. A part of the condensate flows back towards the RPV in counter current to the steam. During the reflux-condenser mode, a counter-current flow limitation (CCFL) must be prevented because this would limit the core cooling. The simulation of CCFL conditions, which is dominated by 3D effects, requires the use of a computational fluid dynamics (CFD) approach. These methods are not yet mature and have to be validated before they can be applied to nuclear reactor safety. Therefore, dedicated experimental data is needed with high resolution in space and time. In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, the “hot leg model” was built at Forschungszentrum Dresden-Rossendorf (FZD). This setup is devoted to optical measurement techniques, therefore, a flat test-section design was chosen with a width of 50 mm. The test-section outlines represent the hot leg of a German Konvoi PWR at a scale of 1:3, which corresponds to a channel height of 250 mm in the straight part of the hot leg. The test-section is mounted between two separators, one simulating the reactor pressure vessel and the other is connected to the steam generator inlet chamber. This allows to perform co-current as well as counter-current flow experiments. Moreover, the hot leg model is built in the pressure vessel of the TOPFLOW facility of FZD, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the vessel. Therefore, the test section can be designed with thin materials and equipped with big size windows like in the hot leg model. The presented air/water experiments focus on the flow structure observed in the region of the riser and of the steam generator inlet chamber at room temperature and pressures up to 3 bars. The performed high-speed observations show the evolution of the stratified interface and the distribution of the two-phase mixture (droplet and bubbles). Counter-current flow limitation, or the onset of flooding, was found by analysing the water levels measured in the separators. A confrontation with the images indicates that the initiation of flooding coincides with the reversal of the flow in the horizontal part of the hot leg due to high air velocities. Afterwards, bigger waves are generated, which develop to slugs. Furthermore, the CCFL data was compared with similar experiments and empirical correlations available in the literature. The agreement of the CCFL curve is good and indicate that the data is relevant for CFD validation purposes. The zero penetration was found at lower values of the Wallis parameter than in most of the previous work, which can be attributed to the rectangular geometry of the hot leg model.

scholarly journals Experimental Characterisation of the Interfacial Structure during Counter-Current Flow Limitation in a Model of the Hot Leg of a PWR

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
Christophe Vallée ◽  
Toshifumi Nariai ◽  
Takashi Futatsugi ◽  
Akio Tomiyama ◽  
Dirk Lucas ◽  
...  
Keyword(s):  
Water Level ◽  
High Speed ◽  
Gas Flow ◽  
Current Flow ◽  
Interfacial Structure ◽  
Flow Limitation ◽  
Two Phase ◽  
Flow Rates ◽  
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, dedicated experiments were performed in a scaled down model ofKobe University. The experiments were performed with air and water at atmospheric pressure and room temperature. At high flow rates, CCFL occurs and the discharge of water to the reactor pressure vessel simulator is limited by the formation of slugs carrying liquid back to the steam generator. The structure of the interface was observed from the side of the channel test section using a high-speed video camera. An algorithm was developed to recognise the stratified interface in the camera frames after background subtraction. This method allows extracting the water level at any position in the image as well as performing further statistical treatments. The evolution of the interfacial structure along the horizontal part of the hot leg is shown by the visualisation of the probability distribution of the water level and analysed in function of the liquid and gas flow rates. The data achieved are useful for the analysis of the flow conditions as well as for the validation of modelling approaches like computational fluid dynamics.

Experimental Study on Model Development for Liquid Accumulation in Steam Generator U-Tube (First Report) Condensing Counter-Current Flow in Single Vertical Pipe

2014 ◽  
Author(s):  
Tatsuya Yamaji ◽  
Kohei Yamazaki ◽  
Yasuo Koizumi ◽  
Hiroyasu Ohtake ◽  
Koji Hasegawa ◽  
...  
Keyword(s):  
Steam Generator ◽  
Current Flow ◽  
Water Film ◽  
Flow State ◽  
Water Drainage ◽  
Two Phase ◽  
Pressurized Water ◽  
Condensed Water ◽  
Counter Current Flow ◽  
Counter Current

Experiments of counter-current two-phase flow of upward steam flow and condensing downward film flow in a pipe were performed. The experiments were intended to examine water accumulation in steam generator U-tubes during intermediate and small break loss-of-coolant accidents of a pressurized water reactor. The inner diameter and the length of a test flow channel used in the experiments were 18 mm and 4 m, respectively. Experiments were performed at higher steam velocity a little than the velocity that was expected just after scram as the first trial. There was no water drainage form the test pipe to the lower plenum. All condensed water was entrained by steam to flow out from the top of the test pipe to the upper plenum. The test pipe was filled with the water lump and the water film, then these were blown up upward and the inner wall of the test pipe became dry. Again the test pipe was filled with the water lump and the water film, then these were blown up upward and the inner wall of the test pipe became dry. This process was iterated at short intervals. The flow state in the test pipe is highly chaotic and agitated. Condensed water flows up and down at high frequencies. It is indicated that to examine the time averaged void fraction and the two-phase pressure drop of the counter-current flow are required.

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