Numerical Investigation of Thermal Stratification in the Primary Circuit of VVER-440 Type Reactors

2006 ◽  
Author(s):  
Ildiko´ Boros ◽  
Attila Aszo´di ◽  
Ga´bor Le´gra´di
Keyword(s):  
Thermal Stratification ◽  
Cooling System ◽  
Maximum Temperature ◽  
Primary Circuit ◽  
Operational Experience ◽  
Extensive Evaluation ◽  
Transient Simulations ◽  
Surge Line ◽  
Core Cooling

Thermal stratification in the primary loops and in the connected pipes can limit the lifetime of the piping, or lead to penetrating cracks due to the stresses caused by the temperature differences and the cyclic temperature changes. Therefore it is essential to determine the thermal hydraulic parameters of the stratified flow. The determination of the affected pipes can be based on the international operational experience and on engineering consideration. The most affected pipes in PWRs are the pressurizer surge line, the injection pipe of the emergency core cooling systems and the feedwater injection pipe of the steam generators. CFD codes can provide an appropriate tool for the examination of the development and the breaking up of the stratification and the determination of the temperature distribution. However, the challenge of the uncertainty of the boundary conditions has to be faced because of the unknown flow circumstances. According to an extensive evaluation, performed in 1998 by the VEIKI, in the VVER-440/213 units of Paks NPP the most affected pipe is the pressurizer surge line [1]. To find out the possible thermal stratification in the surge line, a temperature monitoring system was installed on the YP20 leg of the surge line of the Unit 1 of the Paks NPP in 2000. The measurements showed that during the heat-up period there is a thermal stratification almost all time in the surge line [2]. The maximum temperature differences reach 140 K (140 °C). The surge line has been modeled with the CFD code CFX-5.7. The performed transient simulations confirmed the existence of a thermal stratification in the surge line, but showed permanent recirculation of colder coolant in the lower layer, caused by the asymmetric arrangement of the surge line legs and the asymmetric connection of the two legs to the main loop. In this paper, the surge line model and the results of the transient simulations are presented. The CFD model of the injection pipe of the high pressure Emergency Core Cooling System and the performed simulations for the analysis of occurrence of thermal stratification are presented as well.

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

An Unsteady Analysis on Thermal Stratification in the SCS Piping Branched Off the RCS Piping

2003 ◽  
Author(s):  
Kwang-Chu Kim ◽  
Man-Heung Park ◽  
Hag-Ki Youm ◽  
Sun-Ki Lee ◽  
Tae-Ryong Kim ◽  
...  
Keyword(s):  
Temperature Difference ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Cooling System ◽  
Numerical Study ◽  
Outer Wall ◽  
Maximum Temperature ◽  
Rear Side ◽  
Maximum Temperature Difference ◽  
Starting Point

A numerical study is performed to estimate on an unsteady thermal stratification phenomenon in the Shutdown Cooling System (SCS) piping branched off the Reactor Coolant System (RCS) piping of Nuclear Power Plant. In the results, turbulent penetration reaches to the 1st isolation valve. At 500sec, the maximum temperature difference between top and bottom inner wall in piping is observed at the starting point of horizontal piping passing elbow. The temperature of coolant in the rear side of the 1st isolation valve disk is very slowly increased and the inflection point in temperature difference curve for time is observed at 2700sec. At the beginning of turbulent penetration from RCS piping, the fast inflow generates the higher temperature for the inner wall than the outer wall in the SCS piping. In the case the hot-leg injection piping and the drain piping are connected to the SCS piping, the effect of thermal stratification in the SCS piping is decreased due to an increase of heat loss compared with no connection case. The hot-leg injection piping affected by turbulent penetration from the SCS piping has a severe temperature difference that exceeds criterion temperature stated in reference. But the drain piping located in the rear compared with the hot-leg injection piping shows a tiny temperature difference. In a viewpoint of designer, for the purpose of decreasing the thermal stratification effect, it is necessary to increase the length of vertical piping in the SCS piping, and to move the position of the hot-leg injection piping backward.

scholarly journals CFD Analysis of Passive Autocatalytic Recombiner

2011 ◽  
Vol 2011 ◽  
pp. 1-9
Author(s):  
B. Gera ◽  
P. K. Sharma ◽  
R. K. Singh ◽  
K. K. Vaze
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Maximum Temperature ◽  
Convection Flow ◽  
Heat Of Reaction ◽  
Loss Of Coolant Accident ◽  
Flow Through ◽  
Core Cooling ◽  
Catalytic Recombination ◽  
Steam Heat

In water-cooled nuclear power reactors, significant quantities of hydrogen could be produced following a postulated loss-of-coolant accident (LOCA) along with nonavailability of emergency core cooling system (ECCS). Passive autocatalytic recombiners (PAR) are implemented in the containment of water-cooled power reactors to mitigate the risk of hydrogen combustion. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces below conventional ignition concentration limits and temperature and even in presence of steam. Heat of reaction produces natural convection flow through the enclosure and promotes mixing in the containment. For the assessment of the PAR performance in terms of maximum temperature of catalyst surface and outlet hydrogen concentration an in-house 3D CFD model has been developed. The code has been used to study the mechanism of catalytic recombination and has been tested for two literature-quoted experiments.

Stress and Fatigue Analysis of Pressurizer Surge Line under Thermal Stratification

2019 ◽  
Vol 795 ◽  
pp. 268-275
Author(s):  
Peng Tang ◽  
Zhi Wei Liu ◽  
Hong Wei Qiao ◽  
Peng Zhou Li
Keyword(s):  
Temperature Distribution ◽  
Temperature Difference ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Load Condition ◽  
Maximum Temperature ◽  
Large Temperature ◽  
Maximum Temperature Difference ◽  
Surge Line

Pressurizer surge line is one of the key equipments of nuclear power plants. The thermal stratification due to the intersection of hot and cold fluids inside the pressurizer surge line may affect the safe operation of nuclear power plant. In order to investigate the stress distribution and fatigue characteristics of surge line subjected to long-term thermal stratified loadings, a mechanical model of the surge line was established. And then, according to different temperature distribution assumptions, thermal stress analysis and fatigue assessment were conducted. The results show that the maximum stress appears under the load condition with maximum temperature difference, and finer temperature distribution can obtain more accurate stress and displacement results. The maximum value of fatigue cumulative coefficient appears at the junction of straight pipe and elbow with large temperature difference.

scholarly journals Suggestion for the Proper Treatment of Buoyancy Force Effect on Single Phase Thermal Stratification Phenomenon

2019 ◽  
Vol 128 ◽  
pp. 10009
Author(s):  
Gong-Hee Lee
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Single Phase ◽  
Cooling System ◽  
Turbulent Transport ◽  
Governing Equations ◽  
Service Testing ◽  
Ansys Cfx ◽  
Core Cooling ◽  
Coolant System

In the operating nuclear power plant, piping and components installed in the in-service testing (IST) systems connected to the reactor coolant system are potentially exposed to fatigue caused by thermal stratification flow, possibly threatening its integrity. The emergency core cooling system (ECCS) is one of candidates for which the thermal stratification phenomenon can occur. During the ECCS operating period, buoyancy due to density difference may have significant influence on thermal-hydraulic characteristics of the mixing flow. Therefore, in this study, the need to consider proper buoyancy models in governing equations, especially turbulent transport equations, for accurate prediction of single phase thermal stratification by ECCS injection was numerically studied using ANSYS CFX R17.2.

Optimum Channel Inclination for Gas Venting Under Countercurrent Flow Limitations

2006 ◽  
Author(s):  
Y. Liao ◽  
K. Vierow
Keyword(s):  
Cooling System ◽  
Superficial Velocity ◽  
Countercurrent Flow ◽  
Pressurized Water ◽  
Inclination Angles ◽  
Surge Line ◽  
Water Reactors ◽  
Optimum Channel ◽  
Accident Conditions ◽  
Core Cooling

Countercurrent flow limitation (CCFL) in the pressurizer surge line of future Pressurized Water Reactors (PWR) with passive safety systems is an important phenomenon in reactor safety analysis. The pressurizer surge line is typically comprised of several sections with various inclination angles. Under certain accident conditions, countercurrent flow takes place in the surge line with liquid flowing down from the pressurizer and steam flowing up from the hot leg. The steam venting rate as well as the liquid draining rate may affect the Emergency Core Cooling System (ECCS) actuation. The objective herein is to develop a physics-based model for evaluating the effect of inclination angle on CCFL. For a given liquid superficial velocity in the countercurrent flow system of the pressurizer surge line, the gas superficial velocity should be as large as possible at the onset of flooding, so that the steam can vent as fast as possible without inhibiting the pressurizer drain rate. Thus the system could depressurize in a timely manner to initiate the ECCS actuation. As indicated by CCFL experiments, for a given liquid superficial velocity, the gas superficial velocity attains a greatest value at a certain channel inclination, which is defined as the optimum channel inclination. In the present work, an analytical model is proposed to predict the optimum channel inclination under simplified conditions. The model predictions compare favorably with experimental data.

Numerical Simulation for Effect of Rolling Motion on Thermal Stratification in a Surge Line

2017 ◽  
Author(s):  
Y. L. Shen ◽  
Tao Lu ◽  
Bo Liu
Keyword(s):  
Thermal Stratification ◽  
Fluid Dynamic ◽  
Maximum Temperature ◽  
Rolling Motion ◽  
Fluid Temperature ◽  
Line Pipe ◽  
Static State ◽  
Eddy Simulation ◽  
Surge Line ◽  
Cfd Method

Pressurizer surge lines are essential pipeline structure in NPPs, and the thermal stratification in surge line is recognized as one of the possible cause of thermal fatigue. In this paper, a Computational Fluid Dynamic (CFD) method has been adopted to simulate temperature fluctuations on the process of temperature rising in a pressurizer surge line under rolling motion of single degree of freedom. This work focuses on a fundamental description of differences of thermal stratification between the surge line rolling around the coordinate X-axis condition and that in a static state. The Large-eddy simulation (LES) model is employed to capture the details of temperature change in surge line. Temperature distributions near the inner wall of a surge line pipe with or without swinging were monitored and compared. The temperature differences between the top and bottom of the pipe sections are employed to represent the maximum temperature differences at all the monitored sections. As the surge line swinging, the pattern of temperature distribution and the length of thermal stratification development are different from that in a static. Fluid temperature fluctuation in surge line occur periodically during the fluid temperature rising when the surge line is rotated with the X-axis, and the temperature difference between top and bottom of the surge line is reduced in the same motion mode compared with the static state.

Spray Cooling for Time Resolved Emission Measurements of ICs

2004 ◽  
Author(s):  
Rama R. Goruganthu ◽  
David Bethke ◽  
Shawn McBride ◽  
Tom Crawford ◽  
Jonathan Frank ◽  
...  
Keyword(s):  
Heat Flux ◽  
Thermal Performance ◽  
Flip Chip ◽  
Cooling System ◽  
Spray Cooling ◽  
Junction Temperature ◽  
Maximum Temperature ◽  
Uniform Heat Flux ◽  
High Heat Flux ◽  
Time Resolved

Abstract Spray cooling is implemented on an engineering tool for Time Resolved Emission measurements using a silicon solid immersion lens to achieve high spatial resolution and for probing high heat flux devices. Thermal performance is characterized using a thermal test vehicle consisting of a 4x3 array of cells each with a heater element and a thermal diode to monitor the temperature within the cell. The flip-chip packaged TTV is operated to achieve uniform heat flux across the die. The temperature distribution across the die is measured on the 4x3 grid of the die for various heat loads up to 180 W with corresponding heat flux of 204 W/cm2. Using water as coolant the maximum temperature differential across the die was about 30 °C while keeping the maximum junction temperature below 95 °C and at a heat flux of 200 W/cm2. Details of the thermal performance of spray cooling system as a function of flow rate, coolant

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