scholarly journals Feed Water Line Cracking in Pressurized Water Reactor Plants

10.5772/17247 ◽  
2011 ◽  
Author(s):  
Somnath Chattopadhyay
Keyword(s):  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Line ◽  
Water Reactor

scholarly journals Mode Switching Control for Main Feed-water Process in Pressurized Water Reactor

1999 ◽  
Vol 32 (2) ◽  
pp. 7427-7432
Author(s):  
Futao Zhao ◽  
Wei Du ◽  
Zhiren Hu
Keyword(s):  
Switching Control ◽  
Mode Switching ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Prediction of Transient Hydrodynamic Loads on the PWR Steam Generator Tubes at a Main Feed Water Line Break Accident

2018 ◽  
Author(s):  
Jong Chull Jo ◽  
Jae Jun Jeong ◽  
Byong Jo Yun
Keyword(s):  
Steam Generator ◽  
Weld Line ◽  
Severe Case ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Hydrodynamic Loads ◽  
Water Line ◽  
Steam Generator Tubes ◽  
Present Simulation

A computational fluid dynamics (CFD) analysis was performed to predict the transient hydrodynamic loads exerted on the steam generator tubes and the thrust forces on the broken pipe (which is equal to the impingement forces on target structures in the expanding fluid jet path) during a main feed water line break (FWLB) accident at a pressurized water reactor (PWR) power plant. To address a possible severe case of the transient hydrodynamic loads, the break was assumed to occur at the circumferential weld line between the feed water nozzle and the main feed water pipe so that the compressed sub-cooled water would be discharged through the short broken pipe. Thus, a sub-cooled liquid flashing flow through the broken short feed pipe was simulated numerically. Typical results of the prediction were illustrated and discussed. In addition, the present simulation in terms of the transient mass flow rates during the blowdown following the MFLB was compared to other previous calculations. Based on the discussions, the present simulation is considered to be physically plausible and more realistic than other previous predictions.

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

2010 ◽  
Vol 171-172 ◽  
pp. 379-384
Author(s):  
Khan Salah Ud Din ◽  
Min Jun Peng ◽  
Muhammad Zubair
Keyword(s):  
Heat Removal ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Heat Removal System ◽  
Water Accident ◽  
Relap5 Code ◽  
Removal System ◽  
Residual Heat

In this paper research has been carried out on Loss of Feed Water Accident (LOFW) scenario of the Integral Pressurized Water Reactor ( IPWR) under two circumstances by the use of thermal hydraulic system code i.e Relap5/Mod3.4. In the first one, Passive Residual Heat Removal System (PRHRS) which is designed to absorb core residual heat in case of transient conditions is included which has the function of operating under the accident vulnerabilities. Concerning with the second case i.e without the use of PRHRS rather a tank of water which has the capacity of about 8% of the total feed water supply and is operated under accident scenario is considered. Taken into account these conditions,first the nodalization diagram of the two cases have been figured out then according to the LOFW accident time event scenario use the Relap5 code to simulate the accident. Finally the graphical explanation (separately) of the two cases with graphical approach as well as the conclusion is given at the end.

Analysis of a Feed Water Distribution Ring for a Water Hammer Load Following a Pipe Break

2013 ◽  
Author(s):  
Anders Olsson ◽  
Mikael Möller
Keyword(s):  
Water Distribution ◽  
Pressure Pulse ◽  
Check Valve ◽  
Feed Water ◽  
Piping System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Line ◽  
Load Following ◽  
Pipe Break

In a nuclear pressurized water reactor (PWR) the steam is created in steam generators (SG). The water is returned to the SG via the feed water line and is mixed with the warmer water inside the SG by means of a feed water distribution ring. In the feed water line just outside the SG there is a check valve that will close at pipe break upstream the valve. When the valve closes the water flow through the valve halts and a severe pressure pulse will move into the feed water distribution ring. The pressure pulse is much higher than what the feed water distribution ring can resist statically. However, taking FSI and plasticity into account the magnitude of the pressure pulse is reduced significantly. In this paper an analysis of a feed water distribution ring taking FSI into account with acoustic elements is presented. Without FSI the calculated pressure was 3.7 times higher than the allowable static pressure but taking FSI into account it could be shown that the piping system could be qualified for the event. One of the key issues is the valve model that is elaborated in order to model the closing of the valve as accurately as possible.

Numerical Analysis of Subcooled Water Flashing Flow From a Pressurized Water Reactor Steam Generator Through an Abruptly Broken Main Feed Water Pipe

2019 ◽  
Vol 141 (4) ◽  
Author(s):  
Jong Chull Jo ◽  
Jae Jun Jeong ◽  
Byong Jo Yun ◽  
Jongkap Kim
Keyword(s):  
Flow Field ◽  
Steam Generator ◽  
Feed Water ◽  
Water Pipe ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Subcooled Water ◽  
Numerical Predictions ◽  
Secondary Side

A computational fluid dynamics (CFD) analysis was performed to investigate the hydraulic response of the flow field inside the pressurized water reactor (PWR) steam generator (SG) secondary side and the connected part of main feed water pipe to an abrupt main feed water line break (FWLB) accident. To realistically analyze the transient flow field situation, the flow field was assumed to be occupied initially by highly compressed subcooled water except that the upper part of the SG secondary side where steam occupied as in the practical case and the break was assumed to occur at the circumferential weld line between the feed water nozzle and the main feed water pipe. This would result in a subcooled water flashing flow from the SG through the short-broken pipe end to the surrounding atmosphere, which was numerically simulated in this study. Typical results of the prediction in terms of the fluid transient velocity and pressure were illustrated and discussed. To examine the physical validity of the present numerical simulation of the subcooled water flashing flow, the transient mass flow rates predicted in this study were compared with the other previous numerical predictions based on the subcooled water nonflashing (no phase change) flow or saturated water flashing flow assumptions and the prediction by a simple analysis method.

10 MWe Pressurized Water Reactor for Defense Terrestrial Power Application

1987 ◽  
Author(s):  
W. H. Arnold ◽  
M. C. Gross ◽  
W. S. Shingler ◽  
R. M. Vijuk
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Power Application

Investigation on primary side oriented accident management measures in a hypothetical station blackout scenario for a VVER-1000 pressurized water reactor

Kerntechnik ◽  
2010 ◽  
Vol 75 (1-2) ◽  
pp. 12-19 ◽  
Author(s):  
P. Tusheva ◽  
F. Schäfer ◽  
N. Reinke ◽  
E. Altstadt ◽  
U. Rohde ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Management Measures ◽  
Accident Management ◽  
Station Blackout
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