Heat Removal Performance Evaluation of Several Passive Containment Cooling Systems during Loss of Coolant Accident

1991 ◽  
Vol 28 (10) ◽  
pp. 907-920
Author(s):  
Hirohide OIKAWA ◽  
Hideo NAGASAKA ◽  
Jun-ichiro OTONARI ◽  
Kenji ARAI
Keyword(s):  
Performance Evaluation ◽  
Heat Removal ◽  
Cooling Systems ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
S. P. Saraswat ◽  
P. Munshi ◽  
A. Khanna ◽  
C. Allison
Keyword(s):  
Heat Flux ◽  
Cooling System ◽  
Heat Removal ◽  
Hydraulic Modeling ◽  
Carbon Fiber Composites ◽  
High Heat Flux ◽  
Vacuum Vessel ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Accident Scenarios

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

Recapturing Net Positive Suction Head Margins in Boiling Water Reactor Emergency Core Cooling Systems

2017 ◽  
Author(s):  
Alan J. Bilanin ◽  
Andrew E. Kaufman ◽  
Warren J. Bilanin
Keyword(s):  
Cooling System ◽  
Cooling Water ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Cooling Systems ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Reactor Pressure

Boiling Water Reactor pressure suppression pools have stringent housekeeping requirements, as well as restrictions on amounts and types of insulation and debris that can be present in the containment, to guarantee that suction strainers that allow cooling water to be supplied to the reactor during a Loss of Coolant Accident remain operational. By introducing “good debris” into the cooling water, many of these requirements/restrictions can be relaxed without sacrificing operational readiness of the cooling system.

Development of a Thermal-Hydraulic Code for Chinese Helium-Cooled Ceramic Breeder TBM Cooling System

2013 ◽  
Author(s):  
Jie Wang ◽  
Guanghui Su ◽  
Wenxi Tian ◽  
Suizheng Qiu
Keyword(s):  
Cooling System ◽  
Cooling Systems ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Friction Models ◽  
Cooling Loop ◽  
Thermal Hydraulic Analysis ◽  
Coolant System ◽  
Relap5 Code ◽  
Good Agreement

Helium was chosen as the coolant for divertor cooling loop, Korea helium cooled solid breeder TBM, European helium cooled pebble bed TBM and Chinese helium cooled ceramic breeder TBM. The thermal-hydraulic analysis for the divertor cooling loop and the TBM cooling systems were carried out by RELAP5 and MELOCR codes, which were developed for transient simulation of light water reactor coolant system during postulated accidents. In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system was developed and calculation of the Chinese HCCB TBM cooling system was presented. Heat transfer and flow friction models for helium were added in the code. First, the code was verified by comparing with the RELAP5 code with the same initial and boundary conditions. The first wall temperature, pressure drop and inlet/outlet temperatures were studied and a good agreement was obtained, then ex-vessel loss of coolant accident for Chinese HCCB-TBM cooling system was investigated using TSACO. The results show that the TBM can be cooled efficiently and the TCWS pressure stays within the design limits with a large margin.

Reactor Containment Heat Removal by Passive Heat Sinks Following a Loss-of-Coolant Accident

1977 ◽  
Vol 33 (3) ◽  
pp. 243-247 ◽  
Author(s):  
P. S. Ayyaswamy ◽  
J. N. Chung ◽  
K. K. Niyogi
Keyword(s):  
Heat Removal ◽  
Heat Sinks ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

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.

A model for the analysis of loss of decay heat removal during loss of coolant accident in MTR pool type research reactors

2006 ◽  
Vol 33 (5) ◽  
pp. 405-414 ◽  
Author(s):  
Anis Bousbia-salah ◽  
Brahim Meftah ◽  
Tewfik Hamidouche ◽  
El Khider Si-Ahmed
Keyword(s):  
Heat Removal ◽  
Research Reactors ◽  
Decay Heat ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pool Type

scholarly journals PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
Virpi Kouhia ◽  
Heikki Purhonen ◽  
Vesa Riikonen ◽  
Markku Puustinen ◽  
Riitta Kyrki-Rajamäki ◽  
...  
Keyword(s):  
Steam Generators ◽  
Cooling Systems ◽  
Standard Problem ◽  
Research Activities ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Case Basis ◽  
Reactor Pressure

This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

Thermal Analysis of Severe Channel Damage Caused by a Stagnation Channel Break in a PHWR

2002 ◽  
Vol 124 (2) ◽  
pp. 161-167 ◽  
Author(s):  
D. Mukhopadhyay ◽  
P. Majumdar ◽  
G. Behera ◽  
S. K. Gupta ◽  
V. Venkat Raj
Keyword(s):  
Thermal Analysis ◽  
Heavy Water ◽  
Heat Generation ◽  
Temperature Rise ◽  
Heat Removal ◽  
Reactor Inlet ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Rapid Temperature ◽  
Water Reactors

The reactor channel of the horizontal core of pressurized heavy water reactors experiences very low sustained flow during loss of coolant accident (LOCA) at the reactor inlet feeders caused by certain breaks known as critical channel breaks. In this type of accident the reactor trip is delayed causing a gross mismatch of the heat generation and heat removal in the channel, thus leading to rapid temperature rise in the affected channel. A study has been carried out to identify the phenomena and the break size leading to such a situation. Severe fuel damage is predicted in the channel.

scholarly journals Analysis and Estimation of Core Damage Frequency of Flow Blockage and Loss of Coolant Accident: A Case Study of a 10 MW Water-Water Research Reactor-PSA Level 1

2021 ◽  
Vol 2021 ◽  
pp. 1-17
Author(s):  
F. Ameyaw ◽  
R. Abrefah ◽  
S. Yamoah ◽  
S. Birikorang
Keyword(s):  
Research Reactor ◽  
Heat Removal ◽  
Primary Circuit ◽  
Water Research ◽  
Core Damage ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Level 1 ◽  
Safety Systems ◽  
Flow Blockage

Fault trees (FT) and event trees (ET) are widely used in industry to model and evaluate the reliability of safety systems. This work seeks to analyze and estimate the core damage frequency (CDF) due to flow blockage (FB) and loss of coolant accident (LOCA) due to large rupture of primary circuit pipe with respect to a specific 10 MW Water-Water Research Reactor in Ghana using the FT and ET technique. Using FT, the following reactor safety systems: reactor protection system, primary heat removal system, isolation of the reactor pool, emergency core cooling system (ECCS), natural circulation heat removal, and isolation of the containment were evaluated for their dependability. The probabilistic safety assessment (PSA) Level 1 was conducted using a commercial computational tool, system analysis program for practical coherent reliability assessment (SAPHIRE) 7.0. The frequency of an accident resulting in severe core damage for the internal initiating event was estimated to be 2.51e − 4/yr for the large LOCA as well as 1.45e − 4/yr for FB, culminating in a total core damage frequency of 3.96e − 4/yr. The estimated values for the frequencies of core damage were within the expected margins of 1.0e − 5/yr to 1.0e − 4/yr and of identical sequence of the extent as found for similar reactors.

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