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.

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.

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.

In-Vessel Corium Behavior and Steam Explosion Risks: A Review of Experimental Approaches

2017 ◽  
Author(s):  
Pei Shen ◽  
Wenzhong Zhou
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Radioactive Material ◽  
Severe Accident ◽  
Reactor Accident ◽  
Experimental Approaches

Although no one would like to see, a severe nuclear reactor accident may result in reactor core melting, the fuel melt dropping into water in the reactor vessel, and then interacting with coolant into steam explosion. Steam explosion is a result of very rapid and intense heat transfer and violent interaction between the high temperature melt and low temperature coolant. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. Steam explosion may endanger the reactor vessel and surrounding structures. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the containment integrity. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, fist, small to intermediate scale experiments related to premixing, triggering and propagation stages are reviewed and summarized in tables. Then the intermediate to large scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities’ work is also discussed. The studies on steam explosion are vital for reactor severe accident management, and will lead to improved reactor safety.

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 Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

2012 ◽  
Vol 2012 ◽  
pp. 1-14 ◽  
Author(s):  
H. Yamano ◽  
S. Kubo ◽  
Y. Shimakawa ◽  
K. Fujita ◽  
T. Suzuki ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Natural Circulation ◽  
Heat Removal ◽  
Severe Accident ◽  
Safety Requirements ◽  
Generation Plant ◽  
Safety Design ◽  
Phenomenological Consideration ◽  
Heat Removal System

As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR) adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS) in design extension conditions (DECs). A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs) and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.

Study on Technical Scheme and Frame of Extreme Hazards Mitigation Guideline for CPR1000 NPP

2017 ◽  
Author(s):  
Juanhua Zhang ◽  
Shouyang Zhai ◽  
YeHong Liao
Keyword(s):  
Large Scale ◽  
Command And Control ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Water Storage Tank ◽  
Extreme Hazards ◽  
And Control ◽  
Extensive Damage

The extreme hazards such as large floods, earthquakes, fires or explosion may be lead to an extensive damage or large scale incident in NPP, for instance, loss of command and control, loss of most instruments or equipment which are used for accident mitigation. The typical situations are loss of control room and alternate shutdown capabilities, or loss of AC and DC power, or all of them. Extreme Hazards Mitigation Guideline (EHMG) is developed as the coping strategy for the above extensive damage situations according to the requirements of the characteristic of CPR1000 NPP and the modifications implemented after Fukushima accident. The technical scheme and frame of EHMG includes two modules: EH-IRs (initial response) and EH-MGs (mitigation guideline). And then based on the typical accident analysis, the critical diagnostic parameters and mitigation strategies were made for the EHMG. EH-IRs contains the following contents: Rebuilding the command and control, Off-Site and On-Site Communications, Initial Operational Response Actions, Initial Damage Assessment. EH-MGs contains the following strategies: Manually operating auxiliary turbo pump, Manually depressurizing SGs and injecting into SG, Alternate makeup to RCS, Controlling containment conditions, Alternate Makeup to Spent fuel pool, Alternate Makeup to water storage tank, Containment flooding and so on. The strategies in EHMG can cover both prevention and mitigation phases of severe accident. The first EHMG in China has been validated and implemented in Hong Yan He NPP in September, 2016.

scholarly journals Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

2005 ◽  
Author(s):  
J. Yang ◽  
F. B. Cheung ◽  
J. L. Rempe ◽  
K. Y. Suh ◽  
S. B. Kim
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Boiling Heat Transfer ◽  
Nucleate Boiling ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Integral Effect ◽  
Nucleate Boiling Heat Transfer ◽  
Accident Conditions

Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods.

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