Numerical Two-Phase Flows Simulation and Analysis of the Evolution of the Local Hydrogen Concentration in a PWR Nuclear Containment in the Event of a Severe Accident

2016 ◽  
Author(s):  
Alexandre Zanchetti ◽  
Mickael Hassanaly ◽  
Hervé Cordier ◽  
Antonio Sanna ◽  
Namane Mechitoua ◽  
...  
Keyword(s):  
Hydrogen Concentration ◽  
Nuclear Power ◽  
Cfd Simulation ◽  
Severe Accident ◽  
Water Spray ◽  
Hydrogen Distribution ◽  
Two Phase ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Spray System

The Fukushima accident reminded us of the possible consequences in terms of radiological release that can result from a hydrogen explosion in a nuclear power plant, and, specifically, within the containment of a water cooled reactor building. Some mitigation means against hydrogen hazards exist but performance improvements in numerical tools simulating thermal-hydraulic flows and hydrogen combustion are necessary to allow realistic assessments of severe accident consequences in the containment. In this context, EDF works on CFD simulation of hydrogen distribution in penalized conditions. After dealing with cases for which the water spray system was assumed to be unavailable, and so treated with single-phase CFD code [1] [2], the present paper content is now about simulation and analysis of the local hydrogen concentration in the case of a severe accident for which the water spray system is available. Numerical developments of a multi-phase CFD code (Neptune_CFD) and code validation lead to consistent simulations. The numerical simulation performed by EDF confirms the favorable safety impact of water spray on pressure and temperature for a LOCA scenario occurring on a 1300 MWe Pressurized Water Reactor. Nevertheless, CFD results show that the activation of the spray system before hydrogen injection gives greater hydrogen concentration. So, in the future, to better assess hydrogen risk, EDF will perform computations at CFD taking into account the interaction between combustion and water sprays.

scholarly journals The Fukushima accident was preventable

2015 ◽  
Vol 373 (2053) ◽  
pp. 20140379 ◽  
Author(s):  
Costas Synolakis ◽  
Utku Kânoğlu
Keyword(s):  
Emergency Preparedness ◽  
Nuclear Power ◽  
Hazard Analysis ◽  
Severe Accident ◽  
Regulatory Structure ◽  
Fukushima Accident ◽  
Critical Facilities ◽  
Safety Studies ◽  
New Research ◽  
Large Earthquakes

The 11 March 2011 tsunami was probably the fourth largest in the past 100 years and killed over 15 000 people. The magnitude of the design tsunami triggering earthquake affecting this region of Japan had been grossly underestimated, and the tsunami hit the Fukushima Dai-ichi nuclear power plant (NPP), causing the third most severe accident in an NPP ever. Interestingly, while the Onagawa NPP was also hit by a tsunami of approximately the same height as Dai-ichi, it survived the event ‘remarkably undamaged’. We explain what has been referred to as the cascade of engineering and regulatory failures that led to the Fukushima disaster. One, insufficient attention had been given to evidence of large tsunamis inundating the region earlier, to Japanese research suggestive that large earthquakes could occur anywhere along a subduction zone, and to new research on mega-thrusts since Boxing Day 2004. Two, there were unexplainably different design conditions for NPPs at close distances from each other. Three, the hazard analysis to calculate the maximum probable tsunami at Dai-ichi appeared to have had methodological mistakes, which almost nobody experienced in tsunami engineering would have made. Four, there were substantial inadequacies in the Japan nuclear regulatory structure. The Fukushima accident was preventable, if international best practices and standards had been followed, if there had been international reviews, and had common sense prevailed in the interpretation of pre-existing geological and hydrodynamic findings. Formal standards are needed for evaluating the tsunami vulnerability of NPPs, for specific training of engineers and scientists who perform tsunami computations for emergency preparedness or critical facilities, as well as for regulators who review safety studies.

Study of the Ageing Effects on the Lower Head Failure in a PWR Reactor

2020 ◽  
Author(s):  
R. Lo Frano ◽  
S. Paci ◽  
P. Darnowski ◽  
P. Mazgaj
Keyword(s):  
Nuclear Power ◽  
Quantitative Estimation ◽  
Computer Code ◽  
Severe Accident ◽  
Symmetric Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Term Operation ◽  
Ageing Effects ◽  
Lower Head

Abstract The paper studies influence the ageing effects on the failure of a Reactor Pressure Vessel (RPV) during a severe accident with a core meltdown in a Nuclear Power Plant (NPP). The studied plant is a generic high-power Generation III Pressurized Water Reactor (PWR) developed in the frame of the EU NARSIS project. A Total Station Blackout (SBO) accident was simulated with MELCOR 2.2 severe accident integral computer code. Results of the analysis, temperatures in the lower head and pressures in the lower plenum were used as initial and boundary conditions for the Finite Element Method (FEM) simulations. Two FEM models were developed, a simple two-dimensional axis-symmetric model of the lower head to study fundamental phenomena and complex 3D model to include interactions with the RPV and reactor internals. Ageing effects of a lower head were incorporated into the FEM models to investigate its influence onto lower head response. The ageing phenomena are modelled in terms of degraded mechanical material properties as σ(T), E(T). The primary outcome of the study is the quantitative estimation of the influence of ageing process onto the timing of reactor vessel failure. Presented novel methodology and results can have an impact on future consideration about Long-Term Operation (LTO) of NPPs.

Influence of Gas Properties on Gas-Liquid Two-Phase Flow

2020 ◽  
Author(s):  
Miki Saito ◽  
Taizo Kanai ◽  
Satoshi Nishimura ◽  
Yoshihisa Nishi
Keyword(s):  
Removal Efficiency ◽  
Nuclear Power ◽  
Two Phase Flow ◽  
Wire Mesh ◽  
Severe Accident ◽  
Phase Flow ◽  
Stagnant Water ◽  
Two Phase ◽  
Bubble Characteristics ◽  
Gas Properties

Abstract Understanding the mechanism of fission product (FP) removal by pool scrubbing is essential for improving the prediction accuracy of FP emissions concerning severe accident (SA) in a nuclear power plant. Since FP migrates from a gas-phase to a liquid-phase via a gas-liquid interface, the FP removal efficiency by pool scrubbing is largely affected by the flow regime of gas-liquid two-phase flow. In order to gain a deeper understanding of the influence of gas properties on flow regimes, experiments were performed by injecting helium (He) and nitrogen (N2) gas mixtures of several volumetric ratios through a pool of stagnant water. The result suggests clear effects of gas compositions on gas-liquid two-phase flow, where both void and holdup fractions were found to increase with N2 fraction in the supplied gas. The results were compared with previous studies, and a detailed analysis of bubble characteristics for different compositions of gases was performed using a wire-mesh sensor (WMS). This paper also illustrates further research aspects needed to discuss the effect of its results on FP removal efficiency in a SA, and to acquire comprehensive physics behind such gas property influences on two-phase flow.

Assessment of Radiological Source Term Releases for Potential Severe Accident Scenarios in a PWR SFP

2018 ◽  
Author(s):  
Kwang-Il Ahn ◽  
Jae-Uk Shin
Keyword(s):  
Water Injection ◽  
Severe Accident ◽  
Base Case ◽  
Source Terms ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Makeup Water ◽  
Accident Scenarios ◽  
The Impact

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.

Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile

2016 ◽  
Author(s):  
Yuko Sakamoto ◽  
Koji Shirai ◽  
Toshiko Udagawa ◽  
Shunsuke Kondo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Steel Structures ◽  
Severe Accident ◽  
Numerical Code ◽  
Pressurized Water ◽  
Impact Phenomena ◽  
The Impact

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

scholarly journals Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

2011 ◽  
Vol 2011 ◽  
pp. 1-13 ◽  
Author(s):  
Vadim E. Seleznev ◽  
Vladimir V. Aleshin ◽  
Sergey N. Pryalov
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Dirichlet Boundary ◽  
Dirichlet Boundary Conditions ◽  
Severe Accident ◽  
System Operation ◽  
Spray System ◽  
Severe Accidents ◽  
Full Cross Section ◽  
Reactor Building

The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating) of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.

The Principle and Preliminary Validation of a New Subchannel Code

2013 ◽  
Author(s):  
Ning Bai ◽  
Yuanbing Zhu ◽  
Zhihao Ren ◽  
Haibo He ◽  
Haoliang Lu ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Constitutive Relations ◽  
Reactor Design ◽  
Safety Analysis ◽  
Industrial Applications ◽  
Transfer Model ◽  
Design Codes ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Pressurized Water

Following China’s road map of nuclear technology development, the development of self-reliant nuclear design codes becomes one of the most significant steps in the plan. Among the nuclear design codes, thermal-hydraulic analysis code is indispensable because it is the foundation of reactor safety analysis and reactor design. Recently, China Guangdong Nuclear Power Group has launched a series of R&D projects of reactor design code development. The sub-channel analysis code-LINDEN becomes one of the key subprojects. Since the sub-channel code is developed for thermal-hydraulic design and safety analysis of pressurized water reactors (PWRs), the basic requirements for the LINDEN code are reliability and stability. Therefore, the mathematical model and numerical method developed in the code are based on the matured approaches that have been used in various industrial applications. These models and methods includes: four-equation drift framework model of two-phase flow; the classical heat transfer model and fuel rod model (Poisson equation) as well as the constitutive relations; explicit formulation and stepping algorithms for equation solutions. The solver module of the code is programmed using object-oriented C/C++ language with the structural design.. With all these features, the code was developed to be stable, scalable and compatible. The code’s applicability has been further improved after model improvement and design optimization according to characteristics of China’s proprietary type of reactor. COBRA-IV and LINDEN have been used to conduct the thermal-hydraulics analysis for the Daya bay unit 1 and 2 nuclear plants at the steady-state conditions. The results demonstrate that the two codes agree well with each other. The preliminary tests show that the LINDEN code should be suitable for thermal-hydraulics analysis of large PWRs.

IVR Model Study for Passive PWR

2012 ◽  
Author(s):  
Hong Xu ◽  
Peng Zhang ◽  
Zhiwei Zhou
Keyword(s):  
Nuclear Power ◽  
Safety Evaluation ◽  
Sensitivity Study ◽  
Nuclear Safety ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Plant Design ◽  
Layer Model ◽  
Pressurized Water ◽  
Exterior Surface

1000-MWe scale Pressurized Water Reactor (PWR) is taking service or under construction all over the world, and larger scale plant is studied and developed for its more competitive economics. Not only design basic accidents are analyzed for nuclear safety, the severe accident must also be considered to meet with the increasing requirement of safety. In the “nuclear power plant design safety regulation” (HAF102) issued by Nation Nuclear Safety Administration (NNSA), aim at the preventing and mitigating of severe accident, the regulation bring forward new requirement, which required that during design phase, NPP should consider setting the preventing and mitigation measurement of severe accident as actually as possible. As an approach to prevent the curium from melting down the vessel and entering the containment when a postulated severe accident occurs, In-vessel retention (IVR) of molten core debris via water cooling of the external surface of the reactor vessel has been introduced into AP1000. External reactor vessel cooling (ERVC) is assumed to be achieved keeping exterior surface of vessel at 400K. It is known to all that different scenario and process results in different IVR molten model. As the core melt, different IVR model is formed at different time, such as two-layer model, three-layer model and four layer model. It is necessary to study the IVR model when severe accident process moves on. This paper studies two-layer and three-layer IVR models and find the features of the models. Based on this, sensitivity study of important parameters has also been analyzed. It is useful for us to understand the mechanism of the molten pool. This paper has some directive significance on future IVR strategy research and also provides theoretical support to safety evaluation of PWR plants.

Sign in / Sign up

Export Citation Format

Share Document