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.

Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort. / UOIT

The High Temperature Integrated Multi-Modular Thermal Reactor is a small modular reactor that uses an enhanced conductivity BeO-UO2 fuel with supercritical CO2 coolant to drive turbo-machinery in a direct Brayton cycle. The core consists of several self-contained pressurized modules, each containing fuel elements in pressurized channels surrounded by a graphite moderator, and Brayton cycle turbo-machinery. Each module is subcritical by itself, and when several modules are brought into proximity of one another, a single critical core is formed. The multi-modular approach and use of BeO-UO2 fuel with graphite moderator and supercritical CO2 coolant leads to an inherently safe system capable of high efficiency operation. The pressure channel design and multi-modular approach eliminates engineering challenges associated with large pressure vessels. The subcriticality of the modules ensures inherent safety during construction, transportation, and after decommissioning. Serpent, a continuous-energy Monte-Carlo reactor physics burnup calculation code, was used to develop a critical configuration of the subcritical modules using UO2 fuel enriched with 5 wt% 235U with a 5 wt% BeO additive. The core lifetime was found to be 14.6 years operating at 10 MWth, though the U enrichment and power can be altered to achieve desired core lifetimes. Negative fuel and moderator temperature coefficients of reactivity were found that could maintain safety during operation. The multi-modular design was found to be beneficial compared to a core with all fuel elements in one module. Batch battery type refueling was found to be beneficial and the feasibility of controlling the reactor was demonstrated through the use of control shells that surround each module. The HT-IMMTR design is an inherently safe, highly efficient, economically competitive, and most important, feasible reactor design that takes advantage of proven technologies to facilitate the demonstration of a successful commercial deployment.

A fundamental kinetic model based upon the Hougen-Watson formalism was derived as a basis not only for a better understanding of the reaction behavior but also for the design and simulation of industrial reactors. Kinetic experiments were carried out using a commercial potassium-promoted iron catalyst in a tubular reactor under atmospheric pressure. Typical reaction conditions were temperature = 620oC, steam to ethylbenzene mole ratio = 11, and partial pressure of N2 diluent = 0.432 bar. Experimental data were obtained for different operating conditions, i.e., temperature, feed molar ratio of steam to ethylbenzene, styrene to ethylbenzene, and hydrogen to ethylbenzene and space time. The effluent of the reactor was analyzed on-line using two GCs. Kinetic experiments for the formation of minor by-products, i.e. phenylacetylene, α-methylstyrene, β-methylstyrene, etc, were conducted as well. The reaction conditions were: temperature = 600oC ~ 640oC, a molar ratio of steam to ethylbenzene = 6.5, and partial pressure of N2 diluent = 0.43 bar and 0.64 bar. The products were analyzed by off-line GC. The mathematical model developed for the ethylbenzene dehydrogenation consists of nonlinear simultaneous differential equations in multiple dependent variables. The parameters were estimated from the minimization of the multiresponse objective function which was performed by means of the Marquardt algorithm. All the estimated parameters satisfied the statistical tests and physicochemical criteria. The kinetic model yielded an excellent fit of the experimental data. The intrinsic kinetic parameters were used with the heterogeneous fixed bed reactor model which is explicitly accounting for the diffusional limitations inside the porous catalyst. Multi-bed industrial adiabatic reactors with axial flow and radial flow were simulated and the effect of the operating conditions on the reactor performance was investigated. The dynamic equilibrium coke content was calculated using detailed kinetic model for coke formation and gasification, which was coupled to the kinetic model for the main reactions. The calculation of the dynamic equilibrium coke content provided a crucial guideline for the selection of the steam to ethylbenzene ratio leading to optimum operating conditions.

<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>

Thesis (M.S.)--University of Nevada, Reno, 2007. / "May, 2007." Includes bibliographical references. Online version available on the World Wide Web.

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