Film blowing process is a flexible mass production technology used for manufacturing thin polymeric films. Its flexibility in using an existing die to produce films of different width and thickness, just by controlling process conditions such as, extrudate velocity, excess pressure, and line speed, makes it an attractive process with less capital investment. Controlling the process conditions to obtain a stable bubble, however, is not a trivial task. It is a costly trial and error procedure, which could result is a large wastage of material and other resources. Hence, it is necessary to develop methods to simulate the process and design it using numerical experiments. This important need of the industry defines the objective of this work. In this dissertation, a transient, axisymmetric, nonisothermal, viscoelastic model is developed to simulate the process, and it is solved using finite element method. Material behavior of polymer melt is described using a modified Phan-Thien-Tanner model in the liquid??like region, and anisotropic Kelvin??Voight model in the solid zone, and the transition is modeled using a simple mixture theory. Crystallization kinetics is described using a modified Avrami model with factors to account for the influence of temperature and strain. Results obtained are compared with available experimental results and the model is used to explore stability issues and the role of different parameters. Software developed using this model comes with a GUI based pre- and post-processor, and it can be easily adapted to use other constitutive models.

The proposed work provides a new definition of the pressure derivative function [that is the β-derivative function, Δp βd(t)], which is defined as the derivative of the logarithm of pressure drop data with respect to the logarithm of time This formulation is based on the "power-law" concept. This is not a trivial definition, but rather a definition that provides a unique characterization of "power-law" flow regimes which are uniquely defined by the Δp βd(t) function [that is a constant Δp βd(t) behavior]. The Δp βd(t) function represents a new application of the traditional pressure derivative function, the "power-law" differentiation method (that is computing the dln(Δp)/dln(t) derivative) provides an accurate and consistent mechanism for computing the primary pressure derivative (that is the Cartesian derivative, dΔp/dt) as well as the "Bourdet" well testing derivative [that is the "semilog" derivative, Δpd(t)=dΔp/dln(t)]. The Cartesian and semilog derivatives can be extracted directly from the power-law derivative (and vice-versa) using the definition given above.

A number of mathematical models are used nowadays to describe behaviour of the reallife water distribution system (WDS). It is a well known fact that, to have any meaningful use, any WDS mathematical model must be calibrated first. Here, calibration is defined as process in which a number of WDS model parameters are adjusted until the model mimics behaviour of the real WDS as closely as possible. In this thesis, WDS mathematical models that are used to model water quantity aspect only are analysed. Three hydraulic models considered here are: (1) steady-state flow model, (2) quasi-steady flow (extended period simulation) model and (3) unsteady flow model. The calibration problem analysed here is formulated as a constrained optimisation problem of weighted least square type with the objective defined in a way that enables effective incorporation of prior information on calibration parameters. WDS calibration problem is then analysed in detail, including special issues of identifiability, uniqueness and stability of the problem solution. A list of diagnostic and other statistics and analysis is presented to improve existing calibration approaches by providing partial insight into the calibration process. Calibration of WDS hydraulic models is further improved by the development of new hybrid optimisation method. Being closely related to calibration, the problem of sampling design for calibration of WDS hydraulic models is also addressed here. First, sampling design is formulated as a constrained two-objective optimisation problem. Then, two novel models are developed to solve it. The first model is based on standard, single-objective Genetic Algorithms (SOGA). The second model is based on multi-objective Genetic Algorithms (MOGA). Finally, all novel methodologies presented here are verified successfully on multiple case studies that involve both artificial and real-life WDS. At the end, relevant conclusions are drawn and suggestions for further research work are made.

Shale gas has become increasingly important to United States energy supply. During recent decades, the mechanisms of shale gas storage and transport were gradually recognized. Gas desorption was also realized and quantitatively described. Models and approaches special for estimating rate decline and recovery of shale gas wells were developed. As the strategy of the horizontal well with multiple transverse fractures (MTFHW) was discovered and its significance to economic shale gas production was understood, rate decline and pressure transient analysis models for this type of well were developed to reveal the well behavior. In this thesis, we considered a “Triple-porosity/Dual-permeability” model and performed sensitivity studies to understand long term pressure drawdown behavior of MTFHWs. A key observation from this study is that the early linear flow regime before interfracture interference gives a relationship between summed fracture half-length and permeability, from which we can estimate either when the other is known. We studied the impact of gas desorption on the time when the pressure perturbation caused by production from adjacent transference fractures (fracture interference time) and programmed an empirical method to calculate a time shift that can be used to qualify the gas desorption impact on long term production behavior. We focused on the field case Well A in New Albany Shale. We estimated the EUR for 33 wells, including Well A, using an existing analysis approach. We applied a unified BU-RNP method to process the one-year production/pressure transient data and performed PTA to the resulting virtual constant-rate pressure drawdown. Production analysis was performed meanwhile. Diagnosis plots for PTA and RNP analysis revealed that only the early linear flow regime was visible in the data, and permeability was estimated both from a model match and from the relationship between fracture halflength and permeability. Considering gas desorption, the fracture interference will occur only after several centuries. Based on this result, we recommend a well design strategy to increase the gas recovery factor by decreasing the facture spacing. The higher EUR of Well A compared to the vertical wells encourages drilling more MTFHWs in New Albany Shale.

Braking is mainly controlled by the engine. Friction between a pair of pads and a rotating disc converts the kinetic energy of the vehicle into heat. High temperatures can be reached in the system which can be detrimental for both, components and passenger safety. Numerical techniques help simulate load cases and compute the temperatures field in brake disc and brake pads. The present work implements a Finite Element (FE) toolbox in Matlab/Simulink able to simulate different braking manoeuvres used for brake dimensioning mainly in the early phase of car development process. The brake pad/disc geometry is considered as an axisymmetric body assuming negligible temperature gradient along the circumference of the disc. Calibration using three control factors namely: heat coefficient during braking , acceleration  and emissivity  for the implemented thermal model is performed using experimental investigation at Volvo Car Corporation (VCC) for three specific severe load cases. The thermal model is extended to measure brake fluid temperatures to ensure no vaporisation occurs. Simulation results of the brake disc and brake pad show good correlation with the experimental tests. A sensitivity analysis with the control factors showed convective coefficient during acceleration  the most sensitive, with temperature change of around 16%.

This thesis presents the work involving two broad aspectswithin the field of nuclear reactor analysis and safety. Theseare: - development of a fully independent reactor dynamics andsafety analysis methodology of the RBMK-1500 core transientaccidents and - experiments on the enhancement of coolabilityof a particulate bed or a melt pool due to heat removal throughthe control rod guide tubes. The first part of the thesis focuses on the development ofthe RBMK-1500 analysis methodology based on the CORETRAN codepackage. The second part investigates the issue of coolabilityduring severe accidents in LWR type reactors: the coolabilityof debris bed and melt pool for in- vessel and ex-vesselconditions. The safety of the RBMK type reactors became an importantarea of research after the Chernobyl accident. Since 1989,efforts to adopt Western codes for the RBMK analysis and safetyassessment are being made. The first chapters of this Thesisdescribe the development of an independent neutron dynamics andsafety analysis methodology for the RBMK-1500 core transientsand accidents. This methodology is based on the codes HELIOSand CORETRAN. The RBMK-1500 neutron cross section library wasgenerated with the HELIOS code. The ARROTTA part of theCORETRAN code performs three dimensional neutron dynamicsanalysis and the VIPRE-02 part of the CORETRAN package performsthe rod bundle thermal hydraulics analysis. The VIPRE-02 codewas supplemented with additional CHF correlations, used inRBMK-type reactor calcula tions. The validation, verificationand assessment of the CORETRAN code model for RBMK-1500 wereperformed and are described in the thesis. The second part of the thesis describes the in- vesselparticulate debris bed and melt pool coolabilityinvestigations. The role of the control rod guide tubes (CRGTs)in enhancing the coolability during a postulated severeaccident in a BWR was investigated experimentally. Thisinvestigation is directed towards the accident managementscheme of retaining the core melt within the BWR lowerhead. The particulate debris bed coolability was also investigatedduring the ex-vessel severe accident situation, having a flowof non-condensable gases through the porous debris bed.Experimental investigations on the dependence of the quenchingtime on the non-condensable gas flow rate were carriedout. The first chapter briefly presents the status ofdevelopments in both the RBMK- 1500 core analysis and thecorium coolability areas. The second chapter describes the generation of the RBMK-1500neutron cross section data library with the HELIOS code. Thecross section library was developed for the whole range of thereactor conditions (i.e. for both cold and hot reactor states).The results of the benchmarking with the WIMS-D4 code andvalidation against the RBMK Critical Facility experiments isalso presented here. The HELIOS generated neutron cross sectiondata library provides a close agreement with the WIMS-D4 coderesults. The validation against the data from the CriticalExperiments shows that the HELIOS generated neutron crosssection library provides excellent predictions for thecriticality, axial and radial power distribution, control rodreactivity worths and coolant reactivity effects, etc. Thereactivity effects of voiding for the system, fuel assembly andadditional absorber channel are underpredicted in thecalculations using the HELIOS code generated neutron crosssections. The underprediction, however, is much less than thatobtained when the WIMS-D4 code generated cross sections areemployed. The third chapter describes the work, performed towards theaccurate prediction, assessment and validation of the CHF andpost-CHF heat transfer for the RBMK- 1500 reactor fuelassemblies employing the VIPRE-02 code. This chapter describesthe experiments, which were used for validating the CHFcorrelations, appropriate for the RBMK-1500 type reactors.These correlations after validation were added to the standardversion of the VIPRE-02 code. The VIPRE-02 calculations werebenchmarked against the RELAP5/MOD3.3 code. It was found thatthese user-coded additional CHF correlations developed for theRBMK type reactors (Osmachkin, RRC KI and Khabenskicorrelations) and implemented into the code by the author,provide a good prediction of the CHF occurrence at the RBMKreactor nominal pressure range (at about 7 MPa). Transition andfilm boiling are also predicted well with the VIPRE-02 code forthis pressure range. It was found, that for the RBMK- 1500reactor applications, EPRI CHF correlation should be used forthe CHF predictions for the lower fuel assemblies of thereactor in the subchannel model of the RBMK-1500 fuel assembly.RRC KI and Bowring CHF correlations may be used for the upperfuel assemblies. For a single-channel model of the RBMK-1500fuel channel, Osmachkin, RRC KI and Bowring correlationsprovide the closest predictions and may be used for the CHFestimation. For the low coolant mass fluxes in the fuelchannel, Khabenski correlation can be applied. The fourth chapter presents the verification of the CORETRANcode for the RBMK-1500 core analysis (HELIOS generated neutroncross section data, coupled CORETRAN 3-D neutron kineticscalculations and VIPRE-02 thermal hydraulic module). The modelwas verified against a number of RBMK-1500 plant data andtransient calculations. The new RBMK-1500 core model wassuccessfully applied in several safety assessment applications.A series of transient calculations, considered within the scopeof the RBMK-type reactor Safety Analysis Report (SAR), wereperformed. Several cases of the transient calculations arepresented in this chapter. The HELIOS/CORETRAN/VIPRE-02 coremodel for the RBMK-1500 is fully functional. The RBMK-1500 CPSlogic, added into the CORETRAN provides an adequate response tothe changes in the reactor parameters. Chapters 5 and 6 describe the experiments and the analysisperformed on the coolability of particulate debris bed and meltpool during a postulated severe accident in the LWR. In theChapter 5, the coolability potential, offered by the presenceof a large number of the Control Rod Guide Tubes (CRGTs) in theBWR lower head is presented. The experimental investigationsfor the enhancement of coolability possible with CRGTs wereperformed on two experimental facilities: POMECO (POrous MEdiumCOolability) and COMECO (COrium MElt COolability). Theinfluence of the coolant supply through the CRGT on the debrisbed dryout heat flux, debris bed and melt pool quenching time,crust growth rate, etc. were examined. The heat removalcapacity offered by the presence of the CRGT was quantifiedwith the experimental data, obtained from the POMECO and COMECOfacilities. It was found that the presence of the CRGTs in thelower head of a BWR offers a substantial potential for heatremoval during a postulated severe accident. Additional 10-20kW of heat were removed from the POMECO and COMECO testsections through the CRGT. This corresponds to the average heatflux on the CRGT wall equal to 100-300 kW/m2. In the Chapter 6 the ex-vessel particulate debris bedcoolability is investigated, considering the non-condensablegases released from the concrete ablation process. Theinfluence of the flow of the non-condensable gases on theprocess of quenching a hot porous debris bed was considered.The POMECO test facility was modified, adding the air supply atthe bottom of the test section, to simulate the noncondensablegas release. The process was investigated for both high and lowporosity debris beds. It was found that for the low porositybed composition the countercurrent flooding limit could beexceeded, which would degrade the quenching process for suchbed compositions. The experimental results were analyzed withseveral CCFL models, available in the literature. Keywords:RBMK, light water reactor, core analysis,transient analysis, reactor dynamics, RIA, ATWS, critical heatflux, post-CHF, severe accidents, particulate debris beds, meltpool coolability, BWR, CRGT, dryout, quenching, CCFL, crustgrowth, solidification, water ingression, heat transfer.

Due to the low current ratings of the currently available silicon carbide (SiC) switches they cannot be employed in high-power converters. Thus, it is necessary to parallel-connect several switches in order to reach higher current ratings. This paper presents an investigation of parallel-connected normally-on SiC junction field effect transistors. There are four crucial parameters affecting the effectiveness of the parallel-connected switches. However, the pinch-off voltage and the reverse breakdown voltage of the gates seem to be the most important parameters which affect the switching performance of the devices. In particular, the spread in these two parameters might affect the stable off-state operation of the switches. The switching performance and the switching losses of a pair of parallel-connected devices having different reverse breakdown voltages of the gates is investigated by employing three different gate-driver configurations. It is experimentally shown that using a single gate-driver circuit the switching performance of the parallel-connected devices is almost identical, while the total switching losses are lower compared to the other two configurations. / <p>QC 20121116</p>

The transient analysis of large-scale systems is often difficult even when the systems belong to the simplest M/M/n type of queues. To address analytical difficulties, previous studies have been conducted under various asymptotic regimes by suitably accelerating parameters, thereby establishing some useful mathematical frameworks and giving insights into important characteristics and intuitions. However, some studies show significant limitations when used to approximate real service systems: (i) they are more relevant to steady-state analysis; (ii) they emphasize proofs of convergence results rather than numerical methods to obtain system performance; and (iii) they provide only one set of limit processes regardless of actual system size. Attempting to overcome the drawbacks of previous studies, this dissertation studies the transient analysis of large-scale service systems with time-dependent parameters. The research goal is to develop a methodology that provides accurate approximations based on a technique called uniform acceleration, utilizing the theory of strong approximations. We first investigate and discuss the possible inaccuracy of limit processes obtained from employing the technique. As a solution, we propose adjusted fluid and diffusion limits that are specifically designed to approximate large, finite-sized systems. We find that the adjusted limits significantly improve the quality of approximations and hold asymptotic exactness as well. Several numerical results provide evidence of the effectiveness of the adjusted limits. We study both a call center which is a canonical example of large-scale service systems and an emerging peer-based Internet multimedia service network known as P2P. Based on our findings, we introduce a possible extension to systems which show non-Markovian behavior that is unaddressed by the uniform acceleration technique. We incorporate the denseness of phase-type distributions into the derivation of limit processes. The proposed method offers great potential to accurately approximate performance measures of non-Markovian systems with less computational burden.

The energy wave scattering method (EWS method) makes use of transmission lines and junctions to model the systems, and switches physical quantities to be energy wave variables then simulates the dynamic response of the systems, finally switches the analyzed results from energy wave variables back to physical quantities. Although using EWS method to simulate the dynamic response of structures is still on the initial stage, figuring out the time domain problems as example as transient analysis is suitable for use. Transient analysis is an important segment of dynamic analysis, it needs more extensive mathematics and newer method of calculation. Probably the EWS method is a workable and typical way. The study tries to use the EWS method to simulate the dynamic response of mutildegree of freedom systems, the response are due to different factors such as initial condition factors, damping factors and external force factors else. Let the simulated results display as displacement-time figures and displacement tables, and compared the results from lumped method or the finite element software-ANSYS with system characteristics by the figures and time domain displacements by the tables. On the whole, the simulated results almost matched with the analytical lumped methods. From the results of the study could confirm the feasibility that using the EWS method to simulate the dynamic response of mutildegree of freedom systems, and further tested and verified the applications of the EWS method on the dynamic analysis.

<p>This thesis presents the work involving two broad aspectswithin the field of nuclear reactor analysis and safety. Theseare: - development of a fully independent reactor dynamics andsafety analysis methodology of the RBMK-1500 core transientaccidents and - experiments on the enhancement of coolabilityof a particulate bed or a melt pool due to heat removal throughthe control rod guide tubes.</p><p>The first part of the thesis focuses on the development ofthe RBMK-1500 analysis methodology based on the CORETRAN codepackage. The second part investigates the issue of coolabilityduring severe accidents in LWR type reactors: the coolabilityof debris bed and melt pool for in- vessel and ex-vesselconditions.</p><p>The safety of the RBMK type reactors became an importantarea of research after the Chernobyl accident. Since 1989,efforts to adopt Western codes for the RBMK analysis and safetyassessment are being made. The first chapters of this Thesisdescribe the development of an independent neutron dynamics andsafety analysis methodology for the RBMK-1500 core transientsand accidents. This methodology is based on the codes HELIOSand CORETRAN. The RBMK-1500 neutron cross section library wasgenerated with the HELIOS code. The ARROTTA part of theCORETRAN code performs three dimensional neutron dynamicsanalysis and the VIPRE-02 part of the CORETRAN package performsthe rod bundle thermal hydraulics analysis. The VIPRE-02 codewas supplemented with additional CHF correlations, used inRBMK-type reactor calcula tions. The validation, verificationand assessment of the CORETRAN code model for RBMK-1500 wereperformed and are described in the thesis.</p><p>The second part of the thesis describes the in- vesselparticulate debris bed and melt pool coolabilityinvestigations. The role of the control rod guide tubes (CRGTs)in enhancing the coolability during a postulated severeaccident in a BWR was investigated experimentally. Thisinvestigation is directed towards the accident managementscheme of retaining the core melt within the BWR lowerhead.</p><p>The particulate debris bed coolability was also investigatedduring the ex-vessel severe accident situation, having a flowof non-condensable gases through the porous debris bed.Experimental investigations on the dependence of the quenchingtime on the non-condensable gas flow rate were carriedout.</p><p>The first chapter briefly presents the status ofdevelopments in both the RBMK- 1500 core analysis and thecorium coolability areas.</p><p>The second chapter describes the generation of the RBMK-1500neutron cross section data library with the HELIOS code. Thecross section library was developed for the whole range of thereactor conditions (i.e. for both cold and hot reactor states).The results of the benchmarking with the WIMS-D4 code andvalidation against the RBMK Critical Facility experiments isalso presented here. The HELIOS generated neutron cross sectiondata library provides a close agreement with the WIMS-D4 coderesults. The validation against the data from the CriticalExperiments shows that the HELIOS generated neutron crosssection library provides excellent predictions for thecriticality, axial and radial power distribution, control rodreactivity worths and coolant reactivity effects, etc. Thereactivity effects of voiding for the system, fuel assembly andadditional absorber channel are underpredicted in thecalculations using the HELIOS code generated neutron crosssections. The underprediction, however, is much less than thatobtained when the WIMS-D4 code generated cross sections areemployed.</p><p>The third chapter describes the work, performed towards theaccurate prediction, assessment and validation of the CHF andpost-CHF heat transfer for the RBMK- 1500 reactor fuelassemblies employing the VIPRE-02 code. This chapter describesthe experiments, which were used for validating the CHFcorrelations, appropriate for the RBMK-1500 type reactors.These correlations after validation were added to the standardversion of the VIPRE-02 code. The VIPRE-02 calculations werebenchmarked against the RELAP5/MOD3.3 code. It was found thatthese user-coded additional CHF correlations developed for theRBMK type reactors (Osmachkin, RRC KI and Khabenskicorrelations) and implemented into the code by the author,provide a good prediction of the CHF occurrence at the RBMKreactor nominal pressure range (at about 7 MPa). Transition andfilm boiling are also predicted well with the VIPRE-02 code forthis pressure range. It was found, that for the RBMK- 1500reactor applications, EPRI CHF correlation should be used forthe CHF predictions for the lower fuel assemblies of thereactor in the subchannel model of the RBMK-1500 fuel assembly.RRC KI and Bowring CHF correlations may be used for the upperfuel assemblies. For a single-channel model of the RBMK-1500fuel channel, Osmachkin, RRC KI and Bowring correlationsprovide the closest predictions and may be used for the CHFestimation. For the low coolant mass fluxes in the fuelchannel, Khabenski correlation can be applied.</p><p>The fourth chapter presents the verification of the CORETRANcode for the RBMK-1500 core analysis (HELIOS generated neutroncross section data, coupled CORETRAN 3-D neutron kineticscalculations and VIPRE-02 thermal hydraulic module). The modelwas verified against a number of RBMK-1500 plant data andtransient calculations. The new RBMK-1500 core model wassuccessfully applied in several safety assessment applications.A series of transient calculations, considered within the scopeof the RBMK-type reactor Safety Analysis Report (SAR), wereperformed. Several cases of the transient calculations arepresented in this chapter. The HELIOS/CORETRAN/VIPRE-02 coremodel for the RBMK-1500 is fully functional. The RBMK-1500 CPSlogic, added into the CORETRAN provides an adequate response tothe changes in the reactor parameters.</p><p>Chapters 5 and 6 describe the experiments and the analysisperformed on the coolability of particulate debris bed and meltpool during a postulated severe accident in the LWR. In theChapter 5, the coolability potential, offered by the presenceof a large number of the Control Rod Guide Tubes (CRGTs) in theBWR lower head is presented. The experimental investigationsfor the enhancement of coolability possible with CRGTs wereperformed on two experimental facilities: POMECO (POrous MEdiumCOolability) and COMECO (COrium MElt COolability). Theinfluence of the coolant supply through the CRGT on the debrisbed dryout heat flux, debris bed and melt pool quenching time,crust growth rate, etc. were examined. The heat removalcapacity offered by the presence of the CRGT was quantifiedwith the experimental data, obtained from the POMECO and COMECOfacilities. It was found that the presence of the CRGTs in thelower head of a BWR offers a substantial potential for heatremoval during a postulated severe accident. Additional 10-20kW of heat were removed from the POMECO and COMECO testsections through the CRGT. This corresponds to the average heatflux on the CRGT wall equal to 100-300 kW/m2.</p><p>In the Chapter 6 the ex-vessel particulate debris bedcoolability is investigated, considering the non-condensablegases released from the concrete ablation process. Theinfluence of the flow of the non-condensable gases on theprocess of quenching a hot porous debris bed was considered.The POMECO test facility was modified, adding the air supply atthe bottom of the test section, to simulate the noncondensablegas release. The process was investigated for both high and lowporosity debris beds. It was found that for the low porositybed composition the countercurrent flooding limit could beexceeded, which would degrade the quenching process for suchbed compositions. The experimental results were analyzed withseveral CCFL models, available in the literature.</p><p><b>Keywords:</b>RBMK, light water reactor, core analysis,transient analysis, reactor dynamics, RIA, ATWS, critical heatflux, post-CHF, severe accidents, particulate debris beds, meltpool coolability, BWR, CRGT, dryout, quenching, CCFL, crustgrowth, solidification, water ingression, heat transfer.</p>