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You searched for +publisher:"McMaster University" +contributor:("Luxat, John C."). Showing records 1 – 3 of 3 total matches.

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McMaster University

1. BEHDADI, AZIN. MECHANISTIC MODELLING OF CRITICAL HEAT FLUX ON LARGE DIAMETER TUBES.

Degree: PhD, 2015, McMaster University

Heavy water moderator surrounding each fuel channel is one of the important safety features in CANDU reactors since it provides an in-situ passive heat sink for the fuel in situations where other engineered means of heat removal from fuel channels have failed. In a critical break LOCA scenario, fuel cooling becomes severely degraded due to rapid flow reduction in the affected flow pass of the heat transport system. This can result in pressure tubes experiencing significant heat-up during early stages of the accident when coolant pressure is still high, thereby causing uniform thermal creep strain (ballooning) of the pressure tube (PT) into contact with its calandria tube (CT). The contact of the hot PT with the CT causes rapid redistribution of stored heat from the PT to CT and a large heat flux spike from the CT to the moderator fluid. For conditions where subcooling of the moderator fluid is low, this heat flux spike can cause dryout of the CT. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in continued thermal creep strain deformation of both the PT and the CT. A comprehensive mechanistic model is developed to predict the critical heat flux (CHF) variations along the downward facing outer surface of calandria tube. The model is based on the hydrodynamic model of which considers a liquid macrolayer beneath an elongated vapor slug on the heated surface. Local dryout is postulated to occur whenever the fresh liquid supply to the macrolayer is not sufficient to compensate for the liquid depletion within the macrolayer due to boiling on the heating surface. A boundary layer analysis is performed, treating the two phase motion as an external buoyancy driven flow, to determine the liquid supply rate and the local CHF. The model takes into account different types of flow regime or slip ratio. It is applicable for a calandria vessel as well, under a sever accident condition where a thermal creep failure is postulated to occur if sustained CHF is instigated in the surrounding shield tank water. Model shows good agreement with the available experimental CHF data. The model has been modified to take into account the effect of subcooling and has been validated against the empirical correction factors.

Dissertation

Doctor of Philosophy (PhD)

Advisors/Committee Members: LUXAT, JOHN C., Engineering Physics and Nuclear Engineering.

Subjects/Keywords: CRITICAL HEAT FLUX; TWO PHASE FLOW

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6th Edition):

BEHDADI, A. (2015). MECHANISTIC MODELLING OF CRITICAL HEAT FLUX ON LARGE DIAMETER TUBES. (Doctoral Dissertation). McMaster University. Retrieved from http://hdl.handle.net/11375/18045

Chicago Manual of Style (16th Edition):

BEHDADI, AZIN. “MECHANISTIC MODELLING OF CRITICAL HEAT FLUX ON LARGE DIAMETER TUBES.” 2015. Doctoral Dissertation, McMaster University. Accessed February 16, 2019. http://hdl.handle.net/11375/18045.

MLA Handbook (7th Edition):

BEHDADI, AZIN. “MECHANISTIC MODELLING OF CRITICAL HEAT FLUX ON LARGE DIAMETER TUBES.” 2015. Web. 16 Feb 2019.

Vancouver:

BEHDADI A. MECHANISTIC MODELLING OF CRITICAL HEAT FLUX ON LARGE DIAMETER TUBES. [Internet] [Doctoral dissertation]. McMaster University; 2015. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/11375/18045.

Council of Science Editors:

BEHDADI A. MECHANISTIC MODELLING OF CRITICAL HEAT FLUX ON LARGE DIAMETER TUBES. [Doctoral Dissertation]. McMaster University; 2015. Available from: http://hdl.handle.net/11375/18045


McMaster University

2. Ball, Matthew R. Uncertainty Analysis In Lattice Reactor Physics Calculations.

Degree: PhD, 2011, McMaster University

Comprehensive sensitivity and uncertainty analysis has been performed for light-water reactor and heavy-water reactor lattices using three techniques; adjoint-based sensitivity analysis, Monte Carlo sampling, and direct numerical perturbation. The adjoint analysis was performed using a widely accepted, commercially available code, whereas the Monte Carlo sampling and direct numerical perturbation were performed using new codes that were developed as part of this work. Uncertainties associated with fundamental nuclear data accompany evaluated nuclear data libraries in the form of covariance matrices. As nuclear data are important parameters in reactor physics calculations, any associated uncertainty causes a loss of confidence in the calculation results. The quantification of output uncertainties is necessary to adequately establish safety margins of nuclear facilities. In this work, the propagation of uncertainties associated with both physics parameters (e.g. microscopic cross-sections) and lattice model parameters (e.g. material temperature) have been investigated, and the uncertainty of all relevant lattice calculation outputs, including the neutron multiplication constant and few-group, homogenized cross-sections have been quantified. Sensitivity and uncertainty effects arising from the resonance self-shielding of microscopic cross-sections were addressed using a novel set of resonance integral corrections that are derived from perturbations in their infinite-dilution counterparts. It was found that the covariance of the U238 radiative capture cross-section was the dominant contributor to the uncertainties of lattice properties. Also, the uncertainty associated with the prediction of isotope concentrations during burnup is significant, even when uncertainties of fission yields and decay rates were neglected. Such burnup related uncertainties result solely due to the uncertainty of fission and radiative capture rates that arises from physics parameter covariance. The quantified uncertainties of lattice calculation outputs that are described in this work are suitable for use as input uncertainties to subsequent reactor physics calculations, including reactor core analysis employing neutron diffusion theory.

Doctor of Philosophy (PhD)

Advisors/Committee Members: Novog, David R., Luxat, John C., Engineering Physics.

Subjects/Keywords: nuclear engineering; reactor physics; uncertainty analysis; covariance; cross-sections; resonance self-shielding; Nuclear Engineering; Nuclear Engineering

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6th Edition):

Ball, M. R. (2011). Uncertainty Analysis In Lattice Reactor Physics Calculations. (Doctoral Dissertation). McMaster University. Retrieved from http://hdl.handle.net/11375/11609

Chicago Manual of Style (16th Edition):

Ball, Matthew R. “Uncertainty Analysis In Lattice Reactor Physics Calculations.” 2011. Doctoral Dissertation, McMaster University. Accessed February 16, 2019. http://hdl.handle.net/11375/11609.

MLA Handbook (7th Edition):

Ball, Matthew R. “Uncertainty Analysis In Lattice Reactor Physics Calculations.” 2011. Web. 16 Feb 2019.

Vancouver:

Ball MR. Uncertainty Analysis In Lattice Reactor Physics Calculations. [Internet] [Doctoral dissertation]. McMaster University; 2011. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/11375/11609.

Council of Science Editors:

Ball MR. Uncertainty Analysis In Lattice Reactor Physics Calculations. [Doctoral Dissertation]. McMaster University; 2011. Available from: http://hdl.handle.net/11375/11609


McMaster University

3. Gocmanac, Marko. Critical Heat Flux for a Downwards Facing Disk in a Subcooled Pool Boiling Environment.

Degree: MASc, 2011, McMaster University

An experimental investigation of the physical feasibility of thermal creep failure of the Calandria Vessel under a severe accident load is presented in this thesis. Thermal creep failure is postulated to occur if film boiling is instigated in the Shield Tank Water surrounding the Calandria Vessel. The objective of this experimental study is to measure the Critical Heat Flux (CHF) for a representative geometry in environmental conditions similar to those existing in the CANDU Calandria Vessel and Shield Tank Water. Two geometries of downwards facing surfaces are studied. The first is termed the ‘confined’ study in which bubble motion is demarcated to the heated surface. The second is termed the ‘unconfined’ study where individual bubbles are free to move along the heated surface and vent in any direction. The method used in the confined study is novel and involves the placement of a lip surrounding the heated surface. The level of confinement is adjusted by varying the inclination angle. Data has been obtained for Bond Numbers (Bo) 0, 1.5, 3, 3.6 and 11.8 with corresponding qCHF 596, 495, 295, 223, and 187 kW/m2, respectively. A correlation relating the CHF to level of confinement is stated. The CHF results are in good agreement with Theofanous et. al. (1994), as is the observation that a transition angle is observed in the correlation. The transition angle in this study is found to be ~5.5°. The obtained nucleate boiling curves are compared to Su et. al. (2008) data for similar Bo and excellent agreement is achieved in the medium to high heat flux regions. The unconfined study consists of a downward facing plate in a pool of subcooled water. The obtained nucleate boiling curve is compared with the Stephan-Andelsalam correlation and agreement is not observed. There were visibly different trends in the convective heat transfer coefficient with a mean difference of 31%. The experimental data is compared to data obtained by Nishikawa et. al. (1984) and is found to be in acceptable agreement. The power requirement to instigate film boiling was not met, meaning that the CHF is greater than 1 MW/m2. Visual observations are made and an argument is based on the premise that the phenomenon of dryout for a downwards facing surface is similar to that of an upwards facing surface. The theory and current acceptance of CHF for an upwards facing surface is discussed—in particular Zuber’s “Hydrodynamic Limit” of 1.1 MW/m2, Dhir (1992) and recent experimental evidence from Theofanous et. al. (2002). These three studies were found to be in agreement with results presented here. The experimental evidence presented herein supports the statement that thermal creep failure of the Calandria Vessel is physically unreasonable under analyzed severe accident loads.

Master of Applied Science (MASc)

Advisors/Committee Members: Luxat, John C., Novog, David, Novog, David, Engineering Physics.

Subjects/Keywords: Critical Heat Flux Pool Boiling; Heat Transfer, Combustion; Nuclear Engineering; Heat Transfer, Combustion

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6th Edition):

Gocmanac, M. (2011). Critical Heat Flux for a Downwards Facing Disk in a Subcooled Pool Boiling Environment. (Masters Thesis). McMaster University. Retrieved from http://hdl.handle.net/11375/11813

Chicago Manual of Style (16th Edition):

Gocmanac, Marko. “Critical Heat Flux for a Downwards Facing Disk in a Subcooled Pool Boiling Environment.” 2011. Masters Thesis, McMaster University. Accessed February 16, 2019. http://hdl.handle.net/11375/11813.

MLA Handbook (7th Edition):

Gocmanac, Marko. “Critical Heat Flux for a Downwards Facing Disk in a Subcooled Pool Boiling Environment.” 2011. Web. 16 Feb 2019.

Vancouver:

Gocmanac M. Critical Heat Flux for a Downwards Facing Disk in a Subcooled Pool Boiling Environment. [Internet] [Masters thesis]. McMaster University; 2011. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/11375/11813.

Council of Science Editors:

Gocmanac M. Critical Heat Flux for a Downwards Facing Disk in a Subcooled Pool Boiling Environment. [Masters Thesis]. McMaster University; 2011. Available from: http://hdl.handle.net/11375/11813

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