Research on the set point of feed water system of pressurized water reactor

2015 ◽  
pp. 277-280
Keyword(s):  
Water System ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Set Point ◽  
Water Reactor

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.

Application of 3D Coupled Code SPARKLE-2 to PWR Anticipated Transient Without Scram

2014 ◽  
Author(s):  
Koji Asano ◽  
Hikaru Sakamoto ◽  
Satoshi Imura ◽  
Junto Ogawa
Keyword(s):  
Power Distribution ◽  
Water System ◽  
Nucleate Boiling ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Automatic Reactor ◽  
Main Steam Line ◽  
Main Steam ◽  
Reactor Protection System

An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) followed by failure of the automatic reactor trip function of the reactor protection system. The failure of the reactor to shut down during the certain AOOs can lead to increase in reactor coolant system (RCS) pressure and decrease in departure from nucleate boiling ratio (DNBR) margin for a pressurized water reactor (PWR). Japanese standard PWRs are equipped with ATWS mitigation system which consists of a diverse mitigation system which is independent from the reactor trip system. The ATWS mitigation system automatically initiates isolation of the main steam line flow and the auxiliary feed water system under condition indicative of an ATWS. Mitsubishi Heavy Industries, Ltd. (MHI) applies 3D coupled code, SPARKLE-2 [1] [2], to the ATWS evaluation. SPARKLE-2 is a 3D coupled code developed by MHI and consist of the PWR system transient analysis code M-RELAP5, the 3D neutron kinetics code COSMO-K [3] and the 3D core thermal-hydraulics code MIDAC [4]. SPARKLE-2 implements the 3D characteristics such as local moderator feedback and change in 3D power distribution during transient. Thanks to gain from the 3D calculation, the analysis results show that the plant transients are effectively mitigated by the ATWS mitigation system and the RCS pressure and the minimum DNBR meet the safety criteria. These results also show that operational margins are increased, which enables more flexible design of the reload core.

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.

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