Simulation of nuclear power plant pressurizers with application to an inherently safe reactor.

Persistent Link:
http://hdl.handle.net/10150/184378
Title:
Simulation of nuclear power plant pressurizers with application to an inherently safe reactor.
Author:
Khamis, Ibrahim Ahmad.
Issue Date:
1988
Publisher:
The University of Arizona.
Rights:
Copyright © is held by the author. Digital access to this material is made possible by the University Libraries, University of Arizona. Further transmission, reproduction or presentation (such as public display or performance) of protected items is prohibited except with permission of the author.
Abstract:
Pressurizer modeling for predicting the dynamic pressure of the PIUS system is presented. The transient behavior of this model for the PIUS system was investigated. The validity of this model for the PIUS system is limited to transients that are neither too large nor too long in duration. For example, the model is not capable of describing events following a complete loss of liquid for the pressurizer. However, the model can be used for qualitative prediction of the PIUS system behavior for a wide variety of severe transients. A review of pressurizer modeling indicates that the neglecting of the change in the internal energy of the subcooled water during transients is an acceptable assumption. The inherently safe feature of the PIUS system was confirmed through the self-shutdown of the reactor or, in some cases, through reactor power reduction as a result of the ingress of the pool boric acid solution into the primary system. This dynamic model was constructed of three major components: (1) The primary loop, (2) The secondary loop, and (3) The natural convection loop through the pool. A lumped parameter model, uniform heat transfer, and point kinetics have been the main approximations in this model. Other approximations are mentioned during the modeling of each component of the model. The dynamic model was simulated using the DARE-P continuous system simulation language which was developed in the Electrical Engineering Department at the University of Arizona.
Type:
text; Dissertation-Reproduction (electronic)
Keywords:
Pressurized water reactors -- Simulation methods.; Nuclear power plants -- Safety measures.; Water cooled reactors.
Degree Name:
Ph.D.
Degree Level:
doctoral
Degree Program:
Nuclear and Energy Engineering; Graduate College
Degree Grantor:
University of Arizona

Full metadata record

DC FieldValue Language
dc.language.isoenen_US
dc.titleSimulation of nuclear power plant pressurizers with application to an inherently safe reactor.en_US
dc.creatorKhamis, Ibrahim Ahmad.en_US
dc.contributor.authorKhamis, Ibrahim Ahmad.en_US
dc.date.issued1988en_US
dc.publisherThe University of Arizona.en_US
dc.rightsCopyright © is held by the author. Digital access to this material is made possible by the University Libraries, University of Arizona. Further transmission, reproduction or presentation (such as public display or performance) of protected items is prohibited except with permission of the author.en_US
dc.description.abstractPressurizer modeling for predicting the dynamic pressure of the PIUS system is presented. The transient behavior of this model for the PIUS system was investigated. The validity of this model for the PIUS system is limited to transients that are neither too large nor too long in duration. For example, the model is not capable of describing events following a complete loss of liquid for the pressurizer. However, the model can be used for qualitative prediction of the PIUS system behavior for a wide variety of severe transients. A review of pressurizer modeling indicates that the neglecting of the change in the internal energy of the subcooled water during transients is an acceptable assumption. The inherently safe feature of the PIUS system was confirmed through the self-shutdown of the reactor or, in some cases, through reactor power reduction as a result of the ingress of the pool boric acid solution into the primary system. This dynamic model was constructed of three major components: (1) The primary loop, (2) The secondary loop, and (3) The natural convection loop through the pool. A lumped parameter model, uniform heat transfer, and point kinetics have been the main approximations in this model. Other approximations are mentioned during the modeling of each component of the model. The dynamic model was simulated using the DARE-P continuous system simulation language which was developed in the Electrical Engineering Department at the University of Arizona.en_US
dc.typetexten_US
dc.typeDissertation-Reproduction (electronic)en_US
dc.subjectPressurized water reactors -- Simulation methods.en_US
dc.subjectNuclear power plants -- Safety measures.en_US
dc.subjectWater cooled reactors.en_US
thesis.degree.namePh.D.en_US
thesis.degree.leveldoctoralen_US
thesis.degree.disciplineNuclear and Energy Engineeringen_US
thesis.degree.disciplineGraduate Collegeen_US
thesis.degree.grantorUniversity of Arizonaen_US
dc.identifier.proquest8814247en_US
dc.identifier.oclc701244308en_US
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