I present a stratigraphically layered dual-permeability model composed of thin, alternating, high (~9.2 x 10⁻²⁰ m²) and low (~3.0 x 10⁻²² m²) permeability layers to explain pressure dissipation observed during pulse-decay permeability testing on an intact Barnett Shale core. I combine both layer parallel and layer perpendicular measurements to estimate layer permeability and layer porosity. Micro-computed x-ray tomography and scanning electron microscopy confirm the presence of alternating cm-scale layers of silty-claystone and organic-rich claystone. I interpret that the silty-claystone has a permeability of 9.2 x 10⁻²⁰ m² (92 NanoDarcies) and a porosity of 1.4% and that the organic-rich claystone has a permeability of 3.0 x 10⁻²² m² (0.3 NanoDarcies) and a porosity of 14%. A layered architecture explains the horizontal (k [subscript H] = 107 x 10⁻²¹ m²) to vertical (k [subscript V] = 2.3 x 10⁻²¹ m²) permeability anisotropy ratio observed in the Barnett Shale. These core-scale results suggest that spacing between high-permeability carrier beds can influence resource recovery in shales at the reservoir-scale. I also illustrate the characteristic pulse-decay behavior of core samples with multiple mutually-orthogonal fracture planes, ranging from a single planar fracture to the Warren and Root (1963) "sugar cube" model with three mutually-orthogonal fracture sets. By relating sub core-scale matrix heterogeneity to core-scale gas transport, this work is a step towards upscaling experimental permeability results to describe in-situ gas flow through matrix at the reservoir scale. / text

No description available.

For many projects involving drilled shafts, cores are required to be taken below the shaft base for visual identification of the underlying material. For example, the Texas Department of Transportation (TxDOT) requires a core length of at least 1.5 m (5 ft) or equal to the shaft diameter, whichever is greater, at the shaft base. Although the verification cores are to be extracted at the shaft base, The Department of Transportation of many states do not provide guidance to eliminate the effect of the verification core on the point bearing capacity. A recent study shows that the verification core hole is either filled with concrete in dry condition or with sand-gravel mixture in wet pour (Raibagkar, 2008). This finding is crucial because the point bearing capacity of drilled shafts with an empty hole at the base should be significantly lower than that of drilled shafts without verification core. Although the materials that fill in the verification core remove the risk of losing large point bearing capacity, the exposure of the core holes to air-drying may have an adverse effect on the point bearing capacity tipped in clay shales, especially when the basal material is susceptible to weathering. Therefore, the effect of the verification core on the point bearing capacity has been thoroughly investigated with emphasis on changes in the material properties of four clay shales (Del Rio Clay, Eagle Ford Shale, Taylor Marl, and Navarro Shale) in central Texas. The effect of verification core on the point bearing capacity of drilled shafts was investigated using finite element method (FEM) software, PLAXIS. The results from laboratory tests were converted to input material parameters for Mohr-Coulomb failure criterion, and the thickness of degraded zone around the core was interpreted from fullscale condition degradation tests. The load-displacement curves at the shaft base were created from PLAXIS analyses, and the point bearing capacities were obtained at 5%D and 10%D displacement from load-displacement curves. These capacities were used to calculate reduction factors that relate the point bearing capacity of the reference model (without a verification core) with that of the “core models” (with a verification core). The reduction factors are good indicators to determine if verification core had a positive or negative effect on the point bearing capacity. It was found that the reduction in point bearing capacity of “core models” is typically within 10% capacity of the reference model, and a maximum reduction of 14% was found for the Taylor Marl that was dried for 48 hours. / text

published_or_final_version / Civil Engineering / Master / Master of Philosophy

Understanding pore network structure of a porous medium and fluid flow in the pore network has been an interest to researchers for decades. This study focuses on the characterization and simulation of the pore networks in petroleum reservoir rocks using conventional characterization techniques. A Representative Elemental Volume (REV) model is developed which simulates the pore network as a series of non-interconnected capillary tubes of varying sizes. The model implements mercury porosimetry (MP) results and capillary pressure principles to calculate the size of each bundle of capillary tubes based on a pore throat size distribution produced by the MP experiment. It also implements electrical properties of the rocks to estimate the average length of the capillary tubes. To verify the validity of the simulated network, permeability is calculated for the simulated network using Poiseuille’s flow principles for capillary tubes. Preliminary work showed that the model is capable of simulating the pore network reasonably well because permeability estimations for the simulated network matched measurements. In this study, MP and nuclear magnetic resonance (NMR) tests as well as centrifuge and permeability tests are performed on a suite of 11 sandstone and carbonate rock samples. Because electrical tests were not available, average length of flow paths is calculated with an alternative method that uses porosity to calculate tortuosity. Permeability estimations of the simulated network are compared with measurements. Estimations are also compared to other predictions using methods that implement MP and NMR data to simulate the pore network and the results show that the developed REV model out performs all the other techniques. / October 2015

The research on Determination of Residual Gas Saturation and Gas-Water Relative Permeability in Water-Driven Gas Reservoirs is divided into four stages: literature research, core-flooding experiments, development and application of a new technique for reservoir simulation. Overall, all stages have been completed successfully with several breakthroughs in the areas of Special Core Analysis (SCAL), reservoir engineering and reservoir simulation technology.Initially, a literature research was conducted to survey all available core analysis techniques and their individual characteristics. The survey revealed that there are several core analysis techniques for measuring residual gas saturation (Sgr) and hence, the lack of a commonly agreed method for measuring Sgr. The often-used core analysis techniques are steady-state displacement, co-current imbibition, centrifuge and counter-current imbibition. In this research, all centrifuge tests were performed with a decane-brine system to investigate the possibility of replacing gas with a 'model fluid' to minimise errors due to gas compressibility. Furthermore, Sgr is a function of testing temperature and pressure, types of fluid, wettability, viscosity, flow rate and overburden pressure. Consequently, large uncertainties are associated with measured Sgr and the recoverable gas reserves for water-driven gas reservoirs.Due to the lack of a common method for measuring Sgr, the first important step is to clarify which is the most representative core analysis technique for measuring Sgr. In Stage 2 of the research, core analysis experiments were performed with uniform fluids and ambient temperature. In the core flooding experiments, four different sets of core plugs from various gas reservoirs were selected to cover a wide range of permeability and porosity. Finally, all measured Sgr from the various common core analysis techniques ++ / were compared.The evidence suggested that steady-state displacement and co-current imbibition tests are the most representative techniques for reservoir application. Steady-state displacement also yields the complete relative permeability (RP) data but it requires long stabilisation times and is costly.In the third stage, a new technique was successfully developed for determining both Sgr and gas-water RP data. The new method consists of an initial co-current imbibition experiment followed by the newly developed correlation (Mulyadi, Amin and Kennaird correlation). Co-current imbibition is used to measure the end-point data, for example, initial water saturation (Swi) and Sgr. The MAK correlation was developed to extend the co-current imbibition test by generating gas-water relative permeability data. Unlike previous correlations, MAK correlation is unique because it incorporates and exhibits the formation properties, reservoir conditions and fluid properties (for example, permeability, porosity, interfacial tension and gas density) to generate the RP curves. The accuracy and applicability of MAK correlations were investigated with several sets of gas-water RP data measured by steady-state displacement tests for various gas reservoirs in Australia, New Zealand, South-East Asia and U.S.A. The MAK correlation proved superior to previously developed correlations to demonstrate its robustness.The purpose of the final stage was to aggressively pursue the possibility of advancing the application of the new technique beyond special core analysis (SCAL). As MAK correlation is successful in describing gas water RP in a core plug scale, it is possible to extend its application to describe the overall reservoir flow behaviour. This investigation was achieved by implementing MAK correlation into a 3-D reservoir simulator (MoReS) and performing simulations on a producing ++ / field.The simulation studies were divided into two categories: pre and post upscaled application.The case studies were performed on two X gas-condensate fields: X1 (post upscaled) and X2 (pre upscaled) fields. Since MAK correlation was developed for gas-water systems, several modifications were required to account for the effect of the additional phase (oil) on gas and water RP in gas-condensate systems. In this case, oil RP data was generated by Corey's equations. Five different case studies were performed to investigate the individual and combination effect of implementing MAK correlation, alternative Swi and Sgr correlations and refining porosity and permeability clustering. Moreover, MAK correlation has proven to be effective as an approximation technique for cell by cell simulation to advance reservoir simulation technology.

Thesis (Ph. D.)--University of Texas at Austin, 2008. / Vita. Includes bibliographical references.

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.

Graduation date: 1991

<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>