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.

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.

scholarly journals Aggregation of importance measures for decision making in reliability engineering

2017 ◽  
Vol 231 (3) ◽  
pp. 242-254 ◽  
Author(s):  
Michele Compare ◽  
Michele Bellora ◽  
Enrico Zio
Keyword(s):  
Water System ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Reliability Engineering ◽  
Pressurized Water ◽  
Component Importance ◽  
Majority Principle ◽  
Maintenance Decisions ◽  
Ranking Aggregation

This article investigates the aggregation of rankings based on component importance measures to provide the decision maker with a guidance for design or maintenance decisions. In particular, ranking aggregation algorithms of the literature are considered, a procedure for ensuring that the aggregated ranking is compliant with the Condorcet criterion of majority principle is presented and two original ranking aggregation approaches are proposed. Comparisons are made on a case study of an auxiliary feed-water system of a nuclear pressurized water reactor.

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.

Development of Simulation Codes for Some Accidents in Pressurized Water Reactors

2008 ◽  
Author(s):  
X. B. Yang ◽  
G. H. Su ◽  
S. Z. Qiu
Keyword(s):  
Heat Transfer ◽  
Physical Property ◽  
Friction Model ◽  
Physical Models ◽  
Pressurized Water Reactors ◽  
Transfer Model ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactors

An analysis code has been developed for evaluating the transient thermo-hydraulic behaviors of the pressurized water reactor system. A series of mathematical and physical models is considered in this code, such as the point reactor neutron kinetics model, the heat transfer model, the friction model, the thermo-physical property model and so on. All possible flow and heat transfer conditions in some accidents have been considered and their corresponding models are supplied. Gear’s method is adopted for a better solution to the stiff equations. In this paper, some general accidents in the pressurized water reactors have been investigated, including the station blackout accident (SBO), the loss of flow accident (LOFA), the loss of feed water accident (LOFWA) and the reactivity insertion accident (RIA). The calculated results have been verified by the RELAP5/Mod3 and the results are satisfactory.

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