scholarly journals A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

2016 ◽  
Vol 2 ◽  
pp. 3 ◽  
Author(s):  
Yury Alekseevich Bezrukov ◽  
Vladimir Ivanovitc Schekoldin ◽  
Sergey Ivanovich Zaitsev ◽  
Andrey Nikolaevich Churkin ◽  
Evgeny Aleksandrovich Lisenkov
Keyword(s):  
Reactor Core ◽  
Vver Reactor ◽  
Core Flooding ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

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.

Test and Evaluation of a Filtering System for Retention of Fibres in a Coolant Flow After Loss-of-Coolant Accident

2012 ◽  
Author(s):  
T. Gocht ◽  
W. Kästner ◽  
A. Kratzsch ◽  
M. Strasser
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Time Interval ◽  
Insulation Material ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

In case of an accident the safe heat removal from the reactor core with the installed emergency core cooling system (ECCS) is one of the main features in reactor safety. During a loss-of-coolant accident (LOCA) the release of insulation material fragments in the reactor containment can lead to malfunctions of ECCS. Therefore, the retention of particles by strainers or filtering systems in the ECCS is one of the major tasks. The aim of the presented experimental investigations was the evaluation of a filtering system for the retention of fiber-shaped particles in a fluid flow. The filtering system consists of a filter case with a special lamellar filter unit. The tests were carried out at a test facility with filtering units of different mesh sizes. Insulation material (mineral rock wool) was fragmented to fiber-shaped particles. To simulate the distribution of particle concentration at real plants with large volumes the material was divided into single portions and introduced into the loop with a defined time interval. Material was transported to the filter by the fluid and agglomerated there. The assessment of functionality of the filtering system was made by differential pressure between inlet and outlet of the filtering system and by mass of penetrated particles. It can be concluded that for the tested filtering system no penetration of insulation particles occurred.

Study on the reactor core barrel instantaneous characteristics in case of Loss of Coolant Accident (LOCA) scenarios for loop-type PWR

2017 ◽  
Vol 324 ◽  
pp. 93-102 ◽  
Author(s):  
Mingjun Wang ◽  
Lianfa Wang ◽  
Jianping Jing ◽  
Xinli Gao ◽  
Suizheng Qiu ◽  
...  
Keyword(s):  
Reactor Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

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.

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.

Boiling Water Reactor Core Shroud Acoustic Loads Resulting From a Recirculation Outlet Line Break Loss of Coolant Accident: A Case Study

2011 ◽  
Author(s):  
Daniel Sommerville ◽  
Kumar Karpanan
Keyword(s):  
Reactor Core ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Element Analysis ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Loads Analysis ◽  
Core Shroud

Acoustic loads caused by a postulated Recirculation Line Break Loss of Coolant Accident are one of the required design basis events that must be considered for stress analyses of Boiling Water Reactor internal components such as Jet Pumps and Core Shrouds. These acoustic loads must also be considered for fracture mechanics evaluations performed to assess allowable operating periods for flaws detected during in-service inspections. Acoustic loads methods generally utilized in the past have been 1-D or simplified 2-D models of the domain of interest. In some cases sophisticated nuclear thermal-hydraulic codes are used to obtain the acoustic response to the LOCA event. The present paper summarizes a case study of a RLB LOCA acoustic loads analysis of a BWR core shroud using 3-D linear acoustic finite element analysis. Representative loads are presented at each circumferential shroud weld location. The results presented in this paper can be used to assess the general order of magnitude of the loads which can be expected for similar BWR designs.

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.

scholarly journals Study on Release and Migration of Radionuclides under the Small Break Loss of Coolant Accident in a Marine Reactor

2021 ◽  
Vol 2021 ◽  
pp. 1-12
Author(s):  
Zhao Fang ◽  
Zou Shuliang ◽  
Liu Zejun ◽  
Xu Tao ◽  
Xu Shoulong ◽  
...  
Keyword(s):  
Reactor Core ◽  
Inert Gases ◽  
Hydrogen Release ◽  
Severe Accident ◽  
Release Time ◽  
Primary Circuit ◽  
Analysis Model ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
And Migration

Using severe accident analysis program MELCOR, the small break loss of coolant accident (SBLOCA) analysis model was established for a marine reactor. The release and migration of radionuclides were analyzed during a severe accident induced by SBLOCA. The analysis of the hydrogen source term release showed that the maximum hydrogen release amount was 248.567 kg, and the hydrogen release amount accounted for less than 4% of the air volume. So, there would be no danger of hydrogen explosion accidents. The research mainly focused on the behaviors of the release, the transport, the retention, and the final distribution of inert gases represented by Xe, volatile gases represented by CsI, and nonvolatile nuclides represented by Ba. The results showed that the reactor core exposed completely with a lagging by 510 s and the initial release time of nuclides was lagged by 1916 s. The release shares of Xe in the primary circuit system, the containment, and the environment were 0.013%, 0.06%, and 32.71%, respectively. Also, Ba shared 0.016%, 0.0032%, and 3.28%, respectively, and CsI shared 0.0145%, 0.0012%, and 2.845%, respectively.

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