Numerical and Experimental Investigation of the Behaviour of Non-Contacting Droplets During the Reflood Phase After a LOCA

2010 ◽  
Author(s):  
D. Chatzikyriakou ◽  
S. P. Walker ◽  
C. P. Hale ◽  
D. Lakehal ◽  
G. F. Hewitt
Keyword(s):  
Heat Transfer ◽  
Numerical Simulations ◽  
High Speed ◽  
Secondary Cooling ◽  
Cooling Process ◽  
Hydrodynamic Behaviour ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Wall Interaction ◽  
Transfer Mechanisms

During the reflood phase, following a Large Loss of Coolant Accident (LOCA) in a Pressurised Water Reactor (PWR), a flow of vapour containing small saturated droplets (of order 1mm diameter) is responsible for the precursory cooling before the quenching of the rods by the liquid water. The main mechanism for this cooling process is convective heat transfer to the vapour, with the vapour being cooled by the evaporation of the entrained saturated droplets. If the fuel rod temperature exceeds the Leidenfrost [1] value, the droplets do not wet it, but rather bounce off from it due to the formation of a vapour film between the droplet and the metal. Secondary cooling of the rods is provided by this process. Both the hydrodynamics of these impacts and the droplet-vapour-wall heat transfer mechanisms affect the degree of this secondary cooling. We investigate here the heat transfer attributable to such droplets in typical reflood conditions by a combination of new experimental observations, numerical simulations and correlations based on earlier studies [2], [3], [4]. Using an infrared technique we obtain spatial temperature measurements of the area below a non-contacting droplet [5]. At the same time we observe the hydrodynamic behaviour of the droplet by means of a high speed optical camera. Combining our experimental results with an analytically-computed droplet-wall interaction rate we estimate the cooling by those droplets in typical reflood conditions. These measurements are used for the validation of numerical simulations which are conducted using the CFD code TransAT©, to support its application to cases beyond the present reach of the experimental technique.

The Deformation Rule of the Cooled U75V Heavy Rail Caused by the Different Heat Transfer Coefficient

2011 ◽  
Vol 299-300 ◽  
pp. 1005-1011 ◽  
Author(s):  
Ming Xin Gao ◽  
Pei Long Wang ◽  
Hao Jia ◽  
Shan Hu Tong ◽  
Hua Song ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Transfer Coefficient ◽  
Transfer Coefficient ◽  
High Speed ◽  
Reference Value ◽  
Ansys Software ◽  
Cooling Process ◽  
Technological Parameters ◽  
The Heat Transfer Coefficient ◽  
Heavy Rail

When rolled heavy rail is on the cooling bed for natural cooling, the heat transfer coefficient has important effect on the bending and section sizes of cooled heavy rail. In the paper, the heat-stress couple module ofANSYS software is adopted to carry on numerical simulation on the cooling process of 60kg/m U75V heavy rail, and we obtain the change rule that heat transfer coefficient has effect on bending curvature and section sizes of cooled heavy rail. This study is of great reference value on cooling bed design and the formulation of cooling technological parameters for high speed heavy rail.

Transient Heat Transfer for Helium Gas at Various Flow Decay Time Constants and Heat Generation Rates

2018 ◽  
Author(s):  
Qiusheng Liu ◽  
Ayumi Kitano ◽  
Katsuya Fukuda ◽  
Makoto Shibahara
Keyword(s):  
Heat Transfer ◽  
Heat Generation ◽  
Flow Rate ◽  
Transient Heat Transfer ◽  
Rate Condition ◽  
Initial Flow ◽  
Time Constants ◽  
Loss Of Coolant Accident ◽  
Helium Gas ◽  
Loss Of Coolant

Knowledge of the heat transfer phenomenon under flow decay transient condition is important for the safety assessment of a very high temperature reactor (VHTR) during a loss of coolant accident. In this study, transient heat transfer from a horizontal cylinder to helium gas under exponentially decreasing flow rate condition was experimentally investigated. The experiment was conducted by using a forced convection heat transfer experimental apparatus. A flow control value with its control system was used to realize a flow decay condition. Helium gas was used as a coolant, and a platinum cylinder with a diameter of 1 mm was used as the test heater. A uniform heat generation rate was added to the cylinder by a power source. The cylinder temperature was maintained at an initial value under a definite initial flow rate of the helium gas. Subsequently, the flow rate of the helium gas began to exponentially decrease with different time constants ranging from 3 s to 15 s. The initial flow velocity ranged from 7 m/s to 10 m/s. The surface temperature, heat flux, and heat transfer coefficient were measured during the flow decay transient process under a wide range of experimental conditions such as heat generation rates and flow decay time constants. The results indicated that the temperature of the test heater exhibits a rapid increase during this process, and the increasing rate of the temperature is higher for a lower time constant. An increase in the heat generation rate leads to a higher increase in the surface temperature. Therefore, the heat generation rates of the fuel rods are high when a VHTR operates at high power, and it is more challenging to implement passive safety design to ensure the temperature limitation of the fuel rods during a loss-of-coolant accident. Moreover, the heat transfer coefficient relative to time during the flow rate decreasing process was also obtained. The transient heat transfer process during exponentially decreasing flow rate condition was examined based on the experimental data.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

2006 ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Fluid Temperature ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local Thermal-hydraulic analysis of a Small Break Loss of Coolant Accident transient, the paper presents mainly a parametric study which helps to understand the main phenomena that can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global Thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margin factors.

scholarly journals Small break loss of coolant accident of 200 cm² in cold leg of primary loop of Angra2 nuclear power reactor evaluation

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
EDUARDO MADEIRA BORGES ◽  
GAIANÊ SABUNDJIAN
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Power Reactor ◽  
Reactor Core ◽  
Flow Regimes ◽  
Nuclear Power Reactor ◽  
Primary Loop ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.

Numerical Simulation of the Transient Phase Change Heat Transfer in a Pressurized Water Reactor Core During a Loss-of-Coolant Accident

2011 ◽  
Author(s):  
Soo W. Jo ◽  
Yong K. Lee ◽  
Jong C. Jo
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Transient Heat Transfer ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Temperature of pressurized water reactor (PWR) core is a key parameter used widely for judging the initiation of emergency operating procedures and severe accident management. Since direct measurement of the fuel cladding surface temperature using thermocouples is not practicable currently, the coolant temperature at the core exit locations is monitored instead. Several experimental researches showed that the CET rise during a loss of coolant accident (LOCA) and its magnitudes were always lower than the actual fuel rod cladding temperature at the same time. In this regard, a theoretical analysis of the transient heat transfer of coolant flow in a PWR core is needed to confirm the findings from the previous experimental works. This paper addresses numerical simulation of the transient boiling-induced multiphase flow through a simplified PWR core model during a LOCA by a commercial computational fluid dynamics (CFD) code. The calculated results are discussed to understand the transient heat transfer mechanism in the core and to provide useful technical information for reactor design and operation.

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