An Analysis of Critical Heat Flux in Flow Reversal Transients

1976 ◽  
Vol 98 (2) ◽  
pp. 153-158 ◽  
Author(s):  
R. A. Smith ◽  
F. A. Price ◽  
P. Griffith
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Flow Stagnation ◽  
First Occurrence ◽  
Flow Through ◽  
Flow Reversals

A large inlet break Loss of Coolant Accident in a Pressurized Water Reactor (PWR) would cause the flow through the core to reverse within milliseconds. Currently approved methods of analysis conservatively assume that vapor blanketing of core heat transfer surfaces occurs upon this first reduction to zero flow. A coordinated experimental and analytical study has been conducted to determine when and where the vapor blanketing or Critical Heat Flux (CHF) conditions actually do develop in constant pressure rapid flow reversals. The results indicate that first occurrence of CHF is due not to low coolant velocities, but to flow stagnation in the channel interior with associated rapid channel voiding. Calculations indicate that good cooling should persist over large regions of the core for about 1 s longer than is currently assumed.

scholarly journals Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

2021 ◽  
pp. 4-4
Author(s):  
Salwa Helmy ◽  
Magy Kandil ◽  
Ahmed Refaey
Keyword(s):  
Power Plants ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Safety Systems

In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

Dynamic Assessment of the Core Barrel During Loss of Coolant Accident

2018 ◽  
Author(s):  
Yaroslav Dubyk ◽  
Vladislav Filonov ◽  
Oleksii Ishchenko ◽  
Igor Orynyak ◽  
Yuliia Filonova
Keyword(s):  
Dynamic Assessment ◽  
Pressurized Water Reactor ◽  
Acoustic Response ◽  
Element Analysis ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
The Finite Element Analysis ◽  
Loss Of Coolant ◽  
Duhamel Integral

This article focuses on the dynamic behavior of the Pressurized Water Reactor (PWR) during the Loss Of Coolant Accident (LOCA) which cause the significant acoustic loads on the Core Shrouds. The finite element analysis of a PWR was performed to obtain the acoustic response to the LOCA event. We have performed dynamic stress and strain calculations in the frequency domain for the Core Barrel, according to classical shell theories. The Duhamel integral was used to calculate the transient response of a shell to the transient load caused by the water hammer event. The results obtained were used for fracture mechanics evaluations for flaws, which may occur between inservice inspections.

A Theoretical Critical Heat Flux Model for Rod Bundles Under Pressurized Water Reactor Conditions

1989 ◽  
Vol 85 (2) ◽  
pp. 213-226 ◽  
Author(s):  
Wen-Shan Lin ◽  
Bau-Shei Pei ◽  
Chien-Hsiung Lee ◽  
I. A. Mudawwar
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressurized Water Reactor ◽  
Rod Bundles ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flux Model ◽  
Heat Flux Model

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.

The Thermal-Hydraulic Phenomena Resulting in Early Critical Heat Flux and Rewet in the Semiscale Core

1979 ◽  
Vol 101 (1) ◽  
pp. 43-47 ◽  
Author(s):  
D. M. Snider
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Early Stage ◽  
Flow Reversal ◽  
Core Flow ◽  
Hydraulic Behavior ◽  
The Core ◽  
Flow Channels ◽  
Flow Stagnation ◽  
High Fluid

Analysis was performed to determine the thermal-hydraulic behavior in the electrically heated core simulator of the Semiscale Mod-1 system during the early stage of a simulated LOCA initiated by a large cold leg break. The calculated incore hydraulic behavior was used to obtain a better understanding of early CHF (480 to 700 ms after rupture) and the occurrence of rewet in some locations after CHF. Analysis indicated that shortly after rupture the flow in the upper core stagnated for 600 ms, and the core rapidly voided of coolant. In the center and lower regions of the core, the calculated fluid qualities were between 30 and 70 percent at the time of measured CHF. The high fluid qualities in the flow channels about the heater rods indicated that the mechanism of CHF was dryout of the heater rod surfaces. Critical heat flux did not happen at the location of instantaneous flow stagnation associated with the flow reversal; nor did CHF occur in the region of the prolonged flow stagnation. At about 700 ms after rupture the core flow completely reversed direction, and the influx of coolant from above the heated core was responsible for the measured rewets.

Influence of the CATHARE2 Three-Dimensional Meshing of a 900MWe PWR Core in LOCA Conditions

2008 ◽  
Author(s):  
Valia Guillard ◽  
Nathalie Seiler ◽  
Isabelle Tamburini ◽  
Marie Ducros ◽  
Ste´phanie Massin ◽  
...  
Keyword(s):  
Three Dimensional ◽  
Strong Impact ◽  
Pressurized Water Reactor ◽  
Reference Case ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Lower Head ◽  
3D Meshing

The CATHARE2 code is a “Best-Estimate” system code, developed by the CEA, EDF, AREVA-ANP and IRSN, mainly used in France in the frame of realistic methodology to evaluate safety margins for Pressurized Water Reactor (PWR). Since a three-dimensional thermal hydraulic module, based on a two-fluid 6-equations model, is now available in the most recent versions of the code, a new challenge consists in analyzing whether the 3D modeling of the vessel allows to describe more accurately complex 3D phenomena occurring during Loss of Coolant Accident (LOCA) transients. This document specially studies the sensitivity of the 3D meshing of a 900MWe PWR vessel in Large and Intermediate Breaks LOCA conditions. The development of such a meshing, suitable for the CATHARE2 code, is a long and meticulous task that implies important knowledge of the PWR vessel features and the evaluation of numerous CATHARE2 parameters such as repartition of the guide tubes, volumetric and surfacic porosities or hydraulic diameters. This study also requires a good understanding of the main specific phenomena occurring during the various types of LOCA transients in order to check if the chosen CATHARE2 meshing is well adapted. These studies are successively focused on the consideration of homogenous parameters in the vessel lower head, on axial meshing refinement in the lower head as well as in the active part of the core and finally on radial and azimuthal meshing refinement in the vessel. In order to assess its results and since any experimental evolutions of hydraulics core parameters are available for such multi-dimensional transients, IRSN has considered the physical experimental results of separate effect tests, characteristic of LOCA transients and available in the literature. It has also taken into account the results of previous reference calculations performed with the CATHARE2 code. Considering different three-dimensional meshing of the vessel, the evolution of the hydraulics is observed in the whole reactor and more accurately in the core, which is submitted to a non-uniform radial power profile. In case of Large Break LOCA transient analysis, more attention is paid to the simulation of cross-flows between hot and cold channels during the reflooding phase. This study allows us to verify the strong impact of the axial refinement of the core meshing on the simulation results, leading to greater cross-flows under the quench front. The study of the Intermediate Break LOCA transient, with a delayed pumps stop, has also shown a strong impact of the vessel schematization on the hydraulic evolution. Contrary to the reference case water falling back from hot legs to the upper plenum is observed at the centre of the core and flow is less disturbed in the lower plenum when the CATHARE2 parameters are homogenized. Moreover these sensitivity tests to the 3D meshing show that cladding temperatures are dependent on axial meshing refinement.

Lower Plenum Voiding

1982 ◽  
Vol 104 (3) ◽  
pp. 479-486 ◽  
Author(s):  
D. Bharathan ◽  
G. B. Wallis ◽  
H. J. Richter
Keyword(s):  
Reactor Core ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Coolant Circuit ◽  
Two Phase ◽  
Primary Coolant ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

One of the phenomena involved in a loss-of-coolant accident in a pressurized water reactor may be lower plenum voiding. This might occur during the blowdown phase after a cold-leg break in the primary coolant circuit. Steam generated in the reactor core may flow out of the bottom of the reactor core, turn in the lower plenum of the vessel, in a direction countercurrent to the emergency core coolant flow, and escape via the break. If its velocity is high enough, this steam may sweep water from the bottom (lower plenum) of the reactor vessel. Emergency coolant added to the vessel may also be carried out by the escaping steam and thus the reflooding of the core would be delayed. This paper describes a study of two-phase hydrodynamics associated with lower plenum voiding. Several geometrical configurations were tested at three different scales, using air to simulate the steam. Comparisons were made with data obtained by other researchers.

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