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.

Analysis on Ex-Vessel Steam Explosion in CPR1000+ Unit

2010 ◽  
Author(s):  
Juanhua Zhang ◽  
Jiming Lin ◽  
Shishun Zhang
Keyword(s):  
Physical Model ◽  
Steam Explosion ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Cavity Wall ◽  
Severe Accident ◽  
Severe Accidents ◽  
Structural Capacity ◽  
Impulse Load

Reactor Pit Flooding System (RPF) is adopted under the severe accidents situation in CPR1000+ units. It can move the heat generated from the reactor core via external reactor vessel cooling (ERVC) to keep the integrity of RPV and achieve the in-vessel corium retention (IVR). But if IVR function of RPF is failed, there is Ex-Vessel Steam Explosion (EX-SE) risk. The Ex-Vessel Steam Explosion is analyzed by MC3D software which is for fuel and cooling interaction (FCI). The physical model of CPR1000+ for Steam Explosion is built firstly and then the phenomenon of Ex-Vessel Steam Explosion under typical severe accident is analyzed. The conclusion of this study is that the impulse load of pressure on the cavity wall induced by steam explosion is about 310KPas ∼ 440KPas. Referencing the structure capacity of AP600 containment, if the structural capacity of CPR1000+ containment is equal to AP600, the impulse load of pressure is lower than it. So it could be preliminarily estimated that steam explosion will not threaten the integrality of CPR1000+ containment.

Estimation of Pressure Loads During Ex-Vessel Steam Explosion

2008 ◽  
Author(s):  
Matjazˇ Leskovar
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Phase 1 ◽  
Interaction Process ◽  
Radioactive Material ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central and side melt pours), the cavity water sub-cooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. Special attention was given to melt droplets freezing, which may significantly influence the outcome of the fuel-coolant interaction process. The performed analysis shows that for some ex-vessel steam explosion scenarios much higher pressure loads are predicted than obtained in the OECD program SERENA Phase 1.

Pressure Load Estimation During Ex-Vessel Steam Explosion

2009 ◽  
Vol 131 (3) ◽  
Author(s):  
Matjaž Leskovar
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Phase 1 ◽  
Interaction Process ◽  
Radioactive Material ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor pressure vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel-coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed by varying the location of the melt release (central and side melt pours), the cavity water subcooling, the primary system overpressure at vessel failure, and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. Special attention was given to melt droplet freezing, which may significantly influence the outcome of the fuel-coolant interaction process. The performed analysis shows that for some ex-vessel steam explosion scenarios much higher pressure loads are predicted than obtained in the OECD program SERENA Phase 1.

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

scholarly journals Experimental study of heat transfer through cooling water circuit in a reactor vault by using Al2O3 nano fluid

2018 ◽  
Vol 22 (2) ◽  
pp. 1149-1161 ◽  
Author(s):  
Maria Anish ◽  
Balakrishnan Kanimozh
Keyword(s):  
Heat Transfer ◽  
Temperature Distribution ◽  
Base Fluid ◽  
Nuclear Reactor ◽  
Characteristic Curve ◽  
Reactor Core ◽  
Cooling Water ◽  
Heat Transfer Process ◽  
Nano Fluid ◽  
Al2o3 Nanoparticle

The heat produced in the nuclear reactor due to fission reaction must be kept in control or else it will damage the components in the reactor core. Nuclear plants are using water for the operation dissipation of heat. Instead, some chemical substances which have higher heat transfer coefficient and high thermal conductivity. This experiment aims to find out how efficiently a nanofluid can dissipate heat from the reactor vault. The most commonly used nanofluid is Al2O3 nanoparticle with water or ethylene as base fluid. The Al2O3 has good thermal property and it is easily available. In addition, it can be stabilized in various PH levels. The nanofluid is fed into the reactor?s coolant circuit. The various temperature distribution leads to different characteristic curve that occurs on various valve condition leading to a detailed study on how temperature distribution carries throughout the cooling circuit. As a combination of Al2O3 as a nanoparticle and therminol 55 as base fluid are used for the heat transfer process. The Al2O3 nanoparticle is mixed in therminol 55 at 0.05 vol.% concentration. Numerical analysis on the reactor vault model was carried out by using ABAQUS and the experimental results were compared with numerical results.

scholarly journals Lagrangian Approach for the Study of Heat Transfer in a Nuclear Reactor Core Using the SPH Methodology

2019 ◽  
pp. 108-124
Author(s):  
F. Pahuamba-Valdez ◽  
E. Mayoral-Villa ◽  
C. E. Alvarado-Rodríguez ◽  
J. Klapp ◽  
A. M. Gómez-Torres ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Lagrangian Approach ◽  
Nuclear Reactor Core

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.

Enhancing Probabilistic Risk Assessment Methodology for Solving Reactor Safety Issues

2003 ◽  
Author(s):  
F. L. Cho
Keyword(s):  
Nuclear Reactor ◽  
Severe Accident ◽  
Assessment Methodology ◽  
Reactor Safety ◽  
Reactor Accident ◽  
Safety Issues ◽  
Damage States ◽  
Baseline Information ◽  
Risk Assessment Methodology ◽  
Accident Sequences

This paper reveals a paradigm of analyzing the consequential effects of severe nuclear reactor accident, radionuclides fraction and source terms release, that will influence the MACCS2 codification [1], by coupling with the results of SAPHIA-PSA Levels l & 2 quantification process [2], MELCORE [3], STCP [4], PST [5], and XSOR [6]. Those codes are mutually exclusive and useful. However, it lacks of the closed interface and linkage for addressing Plant Damage States (PDS), Severe Accident Sequences, and Risk Consequence. Thus, it is imperative to formulate the consistent baseline information for MACCS2, PSA Levels 1, 2 and 3, and then linking to a new algorithm of NCM.

Radiation Emergency Readiness Among US Medical Toxicologists: A Survey

2020 ◽  
pp. 1-6 ◽  
Author(s):  
Brian P. Murray ◽  
Eungjae Kim ◽  
Samuel A. Ralston ◽  
Tim P. Moran ◽  
Carol Iddins ◽  
...  
Keyword(s):  
Health Care Providers ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Critical Role ◽  
Medical Toxicology ◽  
Radioactive Material ◽  
Willingness To Participate ◽  
Care Providers ◽  
Cross Sectional Survey ◽  
Cross Sectional

ABSTRACT Introduction: Large scale radiologic and nuclear disasters are rare; however, recent events such as the Fukushima Daiichi nuclear reactor emergency in Japan and current global political tensions have highlighted the need for health-care providers with expertise in managing radiation injuries. Medical Toxicologists have the ability to collaborate with other specialists in filling this critical role. Methods: We conducted a cross-sectional survey to assess the attitudes, experiences, and knowledge of medical toxicologists through the assistance of the American College of Medical Toxicology. Results: The survey was completed by 114 medical toxicologists during the enrollment period. Medical toxicologists who had a willingness to participate in radiologic or nuclear emergencies or who had taken care of patients contaminated with radioactive material were more likely to perform well on the knowledge assessment. Conclusion: We identified that there is a group of medical toxicologists who have the willingness, experience, and knowledge to help manage patients in the event of a radiologic or nuclear emergency.

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