IIST Passive Core Cooling on Pressurizer Top Break

2003 ◽  
Author(s):  
Tay-Jian Liu ◽  
Chien-Hsiung Lee
Keyword(s):  
Cooling System ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Relief Valve ◽  
Pressurized Water ◽  
Hydraulic Behavior ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Core Cooling

Two experiments for small-break loss-of-coolant accident on pressurizer top were conducted at the INER Integral System Test (IIST) facility to investigate thermal-hydraulic behavior of a passive core cooling system (PCCS) in a Westinghouse pressurized water reactor (PWR). The test results are compared with the previous IIST tests under the same initial and boundary conditions for a power-operated relief valve (PORV) stuck-open incident. The objectives of this study are to understand of the key thermal-hydraulic phenomena associated with PCCS and to compare the effectiveness of accident management with or without PCCS. The break sizes were scaled down based on one and all three fully-opened PORVs. This paper identified the key phenomena commonly observed and the phenomena unique to a PWR with PCCS.

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.

Sensitivity Study of LOCA-Generated Debris on the Pressure Drop of Fuel Assembly

2017 ◽  
Author(s):  
Hammad Aslam Bhatti ◽  
Zhangpeng Guo ◽  
Weiqian Zhuo ◽  
Shahroze Ahmed ◽  
Da Wang ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Fuel Assembly ◽  
Cooling System ◽  
Safety Issue ◽  
Cooling Water ◽  
Sensitivity Study ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Core Cooling

The coolant of emergency core cooling system (ECCS), for long-term core cooling (LTCC), comes from the containment sump under the loss-of-coolant accident (LOCA). In the event of LOCA, within the containment of the pressurized water reactor (PWR), thermal insulation of piping and other materials in the vicinity of the break could be dislodged. A fraction of these dislodged insulation and other materials would be transported to the floor of the containment by coolant. Some of these debris might get through strainer and eventually accumulate over the suction sump screens of the emergency core cooling systems (ECCS). So, these debris like fibrous glass, fibrous wool, chemical precipitates and other particles cause pressure drop across the sump screen to increase, affecting the cooling water recirculation. As to address this safety issue, the downstream effect tests were performed over full-scale mock up fuel assembly. Sensitivity studies on pressure drop through LOCA-generated debris, deposited on fuel assembly, were performed to evaluate the effects of debris type and flowrate. Fibrous debris is the most crucial material in terms of causing pressure drop, with fibrous wool (FW) debris being more efficacious than fibrous glass (FG) debris.

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

2014 ◽  
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.

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

2021 ◽  
Vol 134 ◽  
pp. 103648
Author(s):  
Katarzyna Skolik ◽  
Chris Allison ◽  
Judith Hohorst ◽  
Mateusz Malicki ◽  
Marina Perez-Ferragut ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

scholarly journals Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 339-345 ◽  
Author(s):  
Tomasz Bury
Keyword(s):  
Hydrogen Generation ◽  
Cooling System ◽  
Component Mixture ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Issues ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Core Cooling ◽  
Parameter Approach

Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

scholarly journals PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE

2017 ◽  
Vol 19 (2) ◽  
pp. 59 ◽  
Author(s):  
Anhar Riza Antariksawan ◽  
Surip Widodo ◽  
Hendro Tjahjono
Keyword(s):  
Cooling System ◽  
Fuel Temperature ◽  
Blow Down ◽  
Beam Tube ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Dry Out

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.  Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini.  Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

Sign in / Sign up

Export Citation Format

Share Document