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.

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.

Simulation of Loss of Coolant Accident in an Integral Pressurized Water Reactor (IPWR)

2011 ◽  
Vol 230-232 ◽  
pp. 410-414
Author(s):  
Salah Ud Din Khan ◽  
Min Jun Peng ◽  
Muhammad Zubair
Keyword(s):  
Nuclear Power ◽  
Zirconium Hydride ◽  
Pressurized Water Reactor ◽  
Nuclear Power Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Sequence Of Events ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Safety And Reliability

In this paper research has been carried out on the Loss of Coolant Accident (LOCA) in an Integral Pressurized Water Reactor(IPWR) by using thermal hydraulic system code Relap5/Mod3.4.The designing of Integrated Pressurized Water Reactor (IPWR) incorporates the safety and reliability of the reactor to withstand under accidental vulnerabilities. In this study, the reactor under consideration is Uranium Zirconium Hydride Nuclear Power Reactor INSURE-100 with the power output of 100MW.In the current research, the reactor has been described in detail according to the requirement for the simulation of LOCA using Relap5 code with the possibility of occurrence of the time sequence of events. The graphs obtained shows good agreement for the safe operation of IPWR under LOCA.

Transient Analysis for the Case of Changes in Coolant Inventory for Advanced Heavy Water Reactor

2004 ◽  
Author(s):  
H. G. Lele ◽  
A. Srivastava ◽  
B. Chatterjee ◽  
A. J. Gaikwad ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Heavy Water ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Two Phase ◽  
Decay Heat ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Core Cooling

Safety of nuclear reactor needs to be assessed against different categories of Postulated initiating events. Advanced Heavy Water Reactor is natural circulation light water cooled and heavy water moderated pressure tube type of reactor. Inventory of the system is important parameter in determination of flow characteristics of this natural circulation reactor. In view of this, various events that cause changes in PHT system inventory are analysed in this paper. One of the reason for decrease in coolant inventory is hypothetical Loss of coolant accident (LOCA) This event is of very low probability but important from designing engineered safeguard system of a reactor. Loss of coolant accident in a nuclear reactor can cause voiding of the reactor core due to expulsion of primary coolant from break. In such, a situation the reactor core experiences very low heat removal rate from the nuclear fuel though the decay heat generation continues even after tripping of the reactor. Heat generation in the reactor core is due to various sources such as decay heat, stored heat etc, can lead to heating of fuel elements. However, Emergency core cooling systems of the reactor are actuated and prevent undesirable temperature rise. These events are called design basis events and focus is on adequacy of Emergency Core Cooling System (ECCS) and fuel integrity. The scenarios, phenomena encountered and consequences depend upon size and location of break, system characteristics, and actuation and capability of different protection and engineered safeguard systems of the reactor system. Moreover, this reactor has several passive features to ensure safety of this reactor. which are considered in analyzing these events. Events under category of decrease in coolant inventory includes loss of coolant accidents due to break at different locations of different sizes. Various locations considered in this paper are steam line, inlet header, inlet feeder, ECCS header, downcomer, pressure tube, Isolation condenser inlet header, instrument line break at inlet header and steam drum. The paper also considers scenario emerging due to malfunctions like relief valve stuck open. Causes for events under category of increase in coolant inventory are Increase in Drum level controller set point, Inadvertent valving in of Accumulators and Inadvertent valving in of Gravity driven water pool (GDWP). Last two events are not analysed as they are not possible. The analysis for the above events is complex due to various complex and wide ranges of phenomena involved during different pies under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, two-phase natural circulation under depleted inventory conditions. Coupled neutronics and thermal hydraulics behaviour, Phenomena under LOCA, phenomena during ECCS injection, direct injection into fuel rod, advanced accumulator injection., vapour pull through and coupled controller and thermal hydraulics. Modelling of these phenomena for each event is discussed in this paper. In this paper summary of analyses for representive event is presented.

Linear Active Disturbance Rejection Control for Fast Nuclear Reactor

2017 ◽  
Author(s):  
Liu Feng-ming ◽  
Shen Cong ◽  
Zhang Hua-xia ◽  
Zhou Shi-liang ◽  
Liu Yu-yan ◽  
...  
Keyword(s):  
Disturbance Rejection ◽  
Nuclear Power ◽  
Power Reactor ◽  
Power Level ◽  
Reactor Core ◽  
Reactor Power ◽  
Active Disturbance Rejection Control ◽  
Nuclear Power Reactor ◽  
Active Disturbance Rejection ◽  
Disturbance Rejection Control

Since dynamics of a fast nuclear power reactor is quite nonlinear, uncertain and time-varying. Moreover, the reactor period depends on the neutron lifetime and it has a crucial importance in reactor power so it is necessary to have a control system for the reactor power. A reliable method for power-level control named linear active disturbance rejection control (LADRC) based on power-level for a fast nuclear power reactor is presented in this paper. The reactor core is simulated based on the point kinetics equations and one delayed neutron group, fast reactor reactivity equation which is very different from thermal reactors and thermal-hydraulics model of the reactor core. This model is completed by calculating initial value of related parameters and system debugging. Finally, it is prove that this model is reasonable and can be used for controlling. The structure of LADRC is very simple and the Bandwidth-Parameterization based controller tuning is easy. It consists of a PD controller and an extended state observer (ESO). Two LADRC systems are designed for power control with model information and without model information respectively. Simulation results show that both the designed controllers have satisfactory performance over the wide range of reactor operating conditions, while LADRC with model information has faster response, less overshoot, better disturbance rejection ability, greater robustness.

scholarly journals Nuclear Power Reactor Core Melt Accidents

2020 ◽  
Author(s):  
Ahmed Bentaïb ◽  
Hervé Bonneville ◽  
Gérard Cénérino
Keyword(s):  
Nuclear Power ◽  
Power Reactor ◽  
Reactor Core ◽  
Nuclear Power Reactor

An Analysis of a Three-Region Conduction Controlled Rewetting of Hot Surface by Variational Method

2014 ◽  
Author(s):  
Manish K. Agrawal ◽  
Krati Garg ◽  
Aditya Singh ◽  
Santosh K. Sahu
Keyword(s):  
Heat Transfer ◽  
Transfer Rate ◽  
Nuclear Reactor ◽  
The Poor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Solid Liquid Interface ◽  
Solid Liquid ◽  
Hot Surface

When the liquid comes in contact with a sufficiently hot surface, a stable vapor blanket is formed between solid–liquid interface, this significantly reduces the heat transfer rate between liquid and solid interface because of the poor conduction through the vapour layer. As the process continues the surface cools off and the vapor blanket collapses, consequently, the liquid wets the hot surface and this phenomenon is termed as rewetting. The phenomena of rewetting has been observed in many industrial and scientific applications such as emergency core cooling of fuel pins during loss of coolant accident (LOCA) in nuclear reactor, mettalurgical processing and cryogenic chill down during refueling of space transfer vehicles.

SANICRO 71

Alloy Digest ◽  
1965 ◽  
Vol 14 (12) ◽  
Keyword(s):  
Nuclear Power ◽  
Elevated Temperatures ◽  
Power Reactor ◽  
Base Alloy ◽  
Heat Treating ◽  
Good Resistance ◽  
Nuclear Power Reactor ◽  
Temperature Performance ◽  
Resistance To Stress ◽  
Nickel Base

Abstract Sanicro 71 is a nickel-base alloy having good resistance to stress-corrosion, oxidation and creep at elevated temperatures. It is recommended for nuclear power reactor heat exchanger tubes, aircraft turbojet engines and for equipment in the textile, plastic, and chemical industries. This datasheet provides information on composition, physical properties, hardness, elasticity, and tensile properties. It also includes information on high temperature performance and corrosion resistance as well as forming, heat treating, machining, joining, and surface treatment. Filing Code: Ni-108. Producer or source: Sandvik.

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

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