Analysis of mass transfer processes in a reactor during a loss-of-coolant accident

The purpose of the work was to optimize the parameters of the spillage system equipped with a gas pressure hydroaccumulator for a ship pressurized water reactor in a loss-of-coolant accident. The water-gas ratio in the hydroaccumulator and the hydraulic resistance of the path between the hydroaccumulator and the reactor were optimized at the designed hydroaccumulator geometric volume. The main dynamic processes were described using a mathematical model and a computational analysis. A series of numerical calculations were realized to simulate the behavior dynamics of the coolant level in the reactor during the accident – by varying the optimized parameters. Estimates of the minimum and maximum values of the coolant level were obtained: depending on the initial water-gas ratio in the hydroaccumulator at different diameters of the flow restrictor on the path between the hydroaccumulator and the reactor. These results were obtained subject to the restrictive conditions that, during spillage, the coolant level should remain above the core and below the blowdown nozzle. The first condition implies that the core is in safe state, the second excludes the coolant water blowdown. The optimization goal was to achieve the maximum time interval in which these conditions would be satisfied simultaneously. The authors propose methods for selecting the optimal spillage system parameters; these methods provide the maximum time for the core to be in a safe state during a loss-of-coolant accident at the designed hydroaccumulator volume. Using these methods, it is also possible to make assessments from the early stages of designing reactor plants.

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

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp201209004h ◽

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.

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Dynamic Assessment of the Core Barrel During Loss of Coolant Accident

Volume 4: Fluid-Structure Interaction ◽

10.1115/pvp2018-84762 ◽

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.

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

ASME 2011 Pressure Vessels and Piping Conference: Volume 2 ◽

10.1115/pvp2011-57989 ◽

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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Sensitivity Analysis of a PWR Response During a Loss of Coolant Accident Under a Hypothetical Core Blockage Scenario Using RELAP5-3D

Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition ◽

10.1115/icone22-31272 ◽

2014 ◽

Cited By ~ 2

Author(s):

Timothy Crook ◽

Rodolfo Vaghetto ◽

Alessandro Vanni ◽

Yassin A. Hassan

Keyword(s):

Cooling System ◽

Safety Concern ◽

System Response ◽

Flow Paths ◽

Pressurized Water ◽

The Core ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Pass Through ◽

Core Cooling

During a Loss of Coolant Accident (LOCA) a substantial amount of debris may be generated in containment during the blowdown phase. This debris can become a major safety concern since it can potentially impact the Emergency Core Cooling System (ECCS). Debris, produced by the LOCA break flow and transported to the sump, could pass through the filtering systems (debris bed and sump strainer) in the long term cooling phase. If the debris were to sufficiently accumulate at the core inlet region, the core flow could theoretically decrease, affecting the core coolability. Under such conditions, the removal of decay heat would only be possible by coolant flow reaching the core through alternative flow paths, such as the core bypass (baffle). There are certain plant specific features that can play a major role in core cooling from this bypass flow. One of these of key interest is the pressure relief holes. A typical 4-loop Pressurized Water Reactor (PWR) was modeled using RELAP5-3D to simulate the reactor system response during the phases of a large break LOCA and the effectiveness of core cooling under full core blockage was analyzed. The simulation results showed that the presence of alternative flow paths may significantly increase core coolability and prevent cladding temperatures from reaching safety limits, while the lack of LOCA holes may lead to a conservative over-prediction of the cladding temperature.

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Lower Plenum Voiding

Journal of Heat Transfer ◽

10.1115/1.3245118 ◽

1982 ◽

Vol 104 (3) ◽

pp. 479-486 ◽

Cited By ~ 2

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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