Modelling the simplified boiling water reactor natural circulation loop and its stability.

Persistent Link:
http://hdl.handle.net/10150/186405
Title:
Modelling the simplified boiling water reactor natural circulation loop and its stability.
Author:
Latif, Medhat Gamil.
Issue Date:
1993
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:
An integrated model that estimates loop flow rate, heat removal, and stability parameters for the General Electric Simplified Boiling Water Reactor SBWR was developed. The three parameters above used to be calculated individually each by a separate code. The initial approach in loop thermal hydraulic modelling was the steady state solution of the SBWR loop mass, energy, and momentum equations. The power-to-flow map obtained proved to be quite comparable with the Natural Circulation in Boiling Water Reactor (NATBWR) code developed by EPRI, in addition to that of General Electric. At low power levels buoyancy forces are the controlling factor in determining the loop flow rate, while at high power levels two-phase friction losses become the dominating one. Evaluation criteria necessary for comparing different loop geometries performance have been the "minimum critical heat flux ratio (MCHFR)" and the "decay ratio." The predicted flow, from the DFM, at different power levels was used later in a parametric study to answer an important question of which combination of core and riser heights are to be selected that meets both the stability and critical power ratio limits. By modelling bubble time delay through riser in the loop momentum equation, a loop damping coefficient as a measure of loop stability, with higher damping meaning a more stable loop was calculated. Results indicated that during normal operation the SBWR loop is pretty damped. Finally, a detailed code that consists mainly of a fuel pin model, reactor point kinetics for the time dependent reactor normalized power with one group of delayed neutrons, and coolant channel mass, energy, and momentum equations is considered. Reactivity feedbacks from voids and fuel temperature, (Doppler effect), were considered. The loop momentum equation was modified to account for bubble time delay in the riser. After a small perturbation in reactivity, fuel temperature, core average void, and loop flow rate were shown to reach equilibrium values after a period of time equivalent to the transit time of the bubble through the riser. Results from this code matched that of the SBWR safety analysis report.
Type:
text; Dissertation-Reproduction (electronic)
Keywords:
Dissertations, Academic.; Nuclear engineering.
Degree Name:
Ph.D.
Degree Level:
doctoral
Degree Program:
Nuclear and Energy Engineering; Graduate College
Degree Grantor:
University of Arizona
Committee Chair:
Seale, Robert L.

Full metadata record

DC FieldValue Language
dc.language.isoenen_US
dc.titleModelling the simplified boiling water reactor natural circulation loop and its stability.en_US
dc.creatorLatif, Medhat Gamil.en_US
dc.contributor.authorLatif, Medhat Gamil.en_US
dc.date.issued1993en_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.abstractAn integrated model that estimates loop flow rate, heat removal, and stability parameters for the General Electric Simplified Boiling Water Reactor SBWR was developed. The three parameters above used to be calculated individually each by a separate code. The initial approach in loop thermal hydraulic modelling was the steady state solution of the SBWR loop mass, energy, and momentum equations. The power-to-flow map obtained proved to be quite comparable with the Natural Circulation in Boiling Water Reactor (NATBWR) code developed by EPRI, in addition to that of General Electric. At low power levels buoyancy forces are the controlling factor in determining the loop flow rate, while at high power levels two-phase friction losses become the dominating one. Evaluation criteria necessary for comparing different loop geometries performance have been the "minimum critical heat flux ratio (MCHFR)" and the "decay ratio." The predicted flow, from the DFM, at different power levels was used later in a parametric study to answer an important question of which combination of core and riser heights are to be selected that meets both the stability and critical power ratio limits. By modelling bubble time delay through riser in the loop momentum equation, a loop damping coefficient as a measure of loop stability, with higher damping meaning a more stable loop was calculated. Results indicated that during normal operation the SBWR loop is pretty damped. Finally, a detailed code that consists mainly of a fuel pin model, reactor point kinetics for the time dependent reactor normalized power with one group of delayed neutrons, and coolant channel mass, energy, and momentum equations is considered. Reactivity feedbacks from voids and fuel temperature, (Doppler effect), were considered. The loop momentum equation was modified to account for bubble time delay in the riser. After a small perturbation in reactivity, fuel temperature, core average void, and loop flow rate were shown to reach equilibrium values after a period of time equivalent to the transit time of the bubble through the riser. Results from this code matched that of the SBWR safety analysis report.en_US
dc.typetexten_US
dc.typeDissertation-Reproduction (electronic)en_US
dc.subjectDissertations, Academic.en_US
dc.subjectNuclear engineering.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.contributor.chairSeale, Robert L.en_US
dc.contributor.committeememberSecker, Jr., Phillip A.en_US
dc.contributor.committeememberHetrick, David L.en_US
dc.identifier.proquest9408480en_US
dc.identifier.oclc720430700en_US
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