Analysis of Possibility of Pressure Tube Cold Pressurization During ECCS Injection for a Small Break LOCA

2002 ◽  
Vol 124 (4) ◽  
pp. 483-486 ◽  
Author(s):  
D. Mukhopadhyay ◽  
S. K. Gupta ◽  
V. Venkat Raj
Keyword(s):  
Control System ◽  
System Design ◽  
Heavy Water ◽  
Pressure Control ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Control System ◽  
Safe Limits ◽  
Water Reactors

ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.

scholarly journals Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept

2015 ◽  
Vol 2015 ◽  
pp. 1-9
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Severe Accident ◽  
Fusion Reactors ◽  
Transfer System ◽  
Light Water Reactors ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Peak ◽  
Water Reactors ◽  
Heat Transfer System ◽  
Do So

As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters.

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.

Best-Estimate Analysis of Loss-of-Coolant Accident in Pressurized Heavy Water Reactors Using Microcomputers

1992 ◽  
Author(s):  
P. Saha ◽  
T. K. Das ◽  
A. Chanda ◽  
S. Ray
Keyword(s):  
Heavy Water ◽  
Computer Software ◽  
The Other ◽  
Water Reactor ◽  
Canadian French ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Water Reactors

Abstract The present paper discusses the development of a computer software or code for best-estimate analysis of Loss-of-Coolant Accident (LOCA) in Pressurized Heavy Water Reactor (PHWR) systems. The formulation is comparable to U.S., Canadian, French LOCA codes, namely, TRAC, RELAPS, ATHENA, CATHARE, etc. However, the present software has been developed on Microcomputers, namely, PC-XT and AT, whereas the other softwares were developed and are being used primarily on Mainframes such as CDC-7600, CYBER-176, CRAY, etc.

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