Analysis of RESCUE External Reactor Cooling Experiments Using RELAP5 Mod3.3 and ASTEC V2.0 Codes

2013 ◽  
Author(s):  
Martin Kubic ◽  
Milan Bachraty ◽  
Miroslav Barnak ◽  
Peter Matejovic ◽  
David Guenadou ◽  
...  
Keyword(s):  
Heat Flux ◽  
Natural Circulation ◽  
Safety Margin ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Research Network ◽  
Reactor Wall ◽  
Two Phase ◽  
Cooling Loop ◽  
Reactor Surface

Safety margin of the in-vessel retention strategy is given by the difference between thermal load acting on inner reactor surface and coolability limit (in terms of critical heat flux) on outer reactor surface. In order to study the two-phase flow in external reactor vessel cooling loop and heat transfer from curved reactor wall, the RESCUE-2 (Representative loops for External System Cooling Understanding Experiments) experimental facility was erected in CEA Cadarache in France. The facility consists of electrically heated simulator of reactor vessel with an ellipsoidal lower head and cooling loop that enables natural circulation of coolant around the reactor wall. In the frame of SARNET 2 project (Severe Accident Research Network of Excellence, 7th EU Framework Programme) several experiments devoted to external reactor vessel cooling phenomena with certain relevance to VVER-440/V213 reactors, were performed on this facility. The heat flux profile generated by electrical heaters in these experiments was based on the results that were obtained by using ASTEC code for this reactor design. The results of two RESCUE-2 experiments are used in this paper for benchmarking of RELAP 5 Mod.3.3 code and ASTEC V2.0 code.

scholarly journals Studies on Key Effect Factors of Natural Circulation Characteristics for Advanced PWR Reactor Cavity Flooding System

2020 ◽  
Vol 2020 ◽  
pp. 1-11
Author(s):  
Dekui Zhan ◽  
Xinhai Zhao ◽  
Shaoxiong Xia ◽  
Peng Chen ◽  
Huandong Chen
Keyword(s):  
Heat Flux ◽  
Water Level ◽  
Large Scale ◽  
Natural Circulation ◽  
Outer Wall ◽  
Reactor Vessel ◽  
Flow Channel ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Residual Thickness

In order to enhance the ability of severe accident mitigation for Pressurised Water Reactor (PWR), different kinds of severe accident mitigation strategies have been proposed. In-Vessel Retention (IVR) is one of the important severe accident management means by External Reactor Vessel Cooling. Reactor cavity would be submerged to cool the molten corium when a severe accident happens. The success criterion of IVR strategy is that the heat flux which transfers from the corium pool must be lower than the local critical heat flux (CHF) of the reactor pressure vessel (RPV) outside wall and the residual thickness of the RPV wall can maintain the integrity. The residual thickness of RPV is determined by the heat flux transfer from the corium pool and the cooling capability of outer wall of the RPV. There are various factors which would influence the CHF and the cooling capability of outer wall of the RPV. In order to verify the optimized design which is beneficial to the heat transfer and the natural circulation outside the actual reactor vessel, a large-scale Reactor Vessel External Cooling Test (REVECT) facility has been built. A large number of sensitivity tests were carried out, to study how these sensitivity factors affect CHF value and natural circulation. Based on the test results, the structure of the test section flow channel has an obvious effect on the CHF distribution. The flow channel optimized can effectively enhance the CHF value, especially to enhance the CHF value near the “heat focus” region of the molten pool. The water level in the reactor pit has also a great impact on the natural circulation flow. Although natural circulation can be maintained with a low water level, it will lead to a decrease of the cooling capacity. Meanwhile, some noteworthy test phenomena have been found, which are also essential for the design of the reactor pit flooding system.

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.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Loop analysis of a natural circulation two-phase flow under an external reactor vessel cooling

2008 ◽  
Vol 35 (8) ◽  
pp. 1001-1006 ◽  
Author(s):  
Jae Cheol Kim ◽  
Kwang Soon Ha ◽  
Rae Joon Park ◽  
Sang Baik Kim ◽  
Seong Wan Hong
Keyword(s):  
Two Phase Flow ◽  
Natural Circulation ◽  
Reactor Vessel ◽  
Phase Flow ◽  
Loop Analysis ◽  
Two Phase

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.

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