No description available.

The computational ability to accurately simulate boiling water reactor operation under the full range of standard steady-state operation, along with the capability to fully track the isotopic distribution of any fueled region in any location in the core has been developed. This new three-dimensional node-by-node capability can help designers track, for example, a full suite of minor and major actinides, fission products, and even light elements that result from depletion, decay, or transmutations. This isotopic tracking capability is not restricted to BWRs and can be employed in the modeling of PWRs, CANDUs, and other reactor types that can be modeled with the NESTLE code, the base core simulator employed in this research. To accurately simulate boiling water reactor operation, a major thermal-hydraulics upgrade was performed which involved the implementation of a drift-flux solution scheme to model steady-state boiling water flow. Sub-cooled boiling and bulk boiling are accurately modeled and a scheme for computing the correct flow distribution has been implemented. In addition, the incorporation of a nodal ORIGEN-based microscopic depletion solution has been included which allows for exceptional detail in tracking a large number of elements in every node of a core design, thus accounting for spectral dependencies such as moderator density effects, moderator temperature effects, fuel temperature effects, as well as controlled or uncontrolled conditions. The results of this study show the excellent fidelity of the two-phase solution for accurately predicting the boiling of water when compared to experimental results. Likewise, the isotopic inventory results show near-identical agreement with the well-established and validated ORIGEN-based SCALE/TRITON isotopic depletion sequence. The aim of these developments is to eventually produce a publicly available three-dimensional core simulator capable of assessing detailed isotopic inventories, a capability particularly valuable for the evaluation of recycling scenarios and actinide management in a variety of reactor types and fuel designs.

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.

<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>

The ability of coated particles of enriched uranium dioxide (UO2) fuel to withstand high temperatures and contain the fission products in the case of a loss of cooling event is a vital passive safety measure over traditional nuclear fuel requiring active safety systems to provide cooling. As a possible solution towards enhancing the safety of light-water reactors (LWRs), it is envisaged that the fuel in the form of loose-coated particles in a helium atmosphere can be introduced inside Silicon-Carbide nuclear reactor fuel cladding tubes of the fuel elements. The coated particles in this investigation were treated as a bed from where heat was transferred to the cladding tube by means of helium gas and the gas movement was by natural convection. Hence, it is proposed that light-water reactors (LWR) could be made safer by redesigning the fuel in the fuel assembly (see Fig. 1.3b). As a first step towards the implementation of this proposal, a proper understanding of the mechanisms of heat transfer, fluid flow and pressure drop through a packed bed of spheres during natural convection was of utmost importance. Such an understanding was achieved through a review of existing literature on porous media. However, most heat transfer correlations and models in heated packed beds are for forced convectional conditions and as such characterise porous media as a function of Reynolds number only rather than expressing media heat transfer performance as a function of thermal properties of the bed in combination with the various components of the overall heat transfer. The media heat transfer performance considered as a function of thermal properties of the bed in the proposed design is found to be a more appropriate approach than the media as a function of Reynolds number. The quest to examine the particle-to-fluid heat transfer characteristics expected in the proposed new fuel design led to implementing this research work in three phases, namely experimental, theoretical and numerical simulation. An experimental investigation of fluid-to-particle natural convection heat transfer characteristics in packed beds heated from below was carried out. Captured data readings from the experiment were analysed and heat transfer characteristics in the medium evaluated by applying the first principle heat transfer concept. A basic unit cell (BUC) model was developed for the theoretical analysis and applied to determine the heat transfer coefficient, h, of the medium. The model adopted a concept in which a single unit of the packed bed was analysed and taken as representative of the entire bed; it related the convective heat transfer effect of the flowing fluid with the conduction and radiative effect at the finite contact spot between adjacent unit cell particles. As a result, the model could account for the thermophysical properties of sphere particles and the heated gas, the interstitial gas effect, gas temperature, contact interface between particles, particle size and particle temperature distribution in the investigated medium. Although the heat transfer phenomenon experienced in the experimental set-up was a reverse case of the proposed fuel design, the study with the achievement in the validation with the Gunn correlation aided in developing the appropriate theoretical relations required for evaluating the heat transfer characteristics in the proposed nuclear fuel design. A slender geometrical model mimicking the proposed nuclear fuel in the cladding was numerically simulated to investigate the heat transfer characteristics and flow distribution under the natural convective conditions anticipated in beds of randomly packed spheres (coated fuel particles) using a commercial code. Random packing of the particles was achieved by discrete element method (DEM) simulation with the aid of Star CCM+ while particle-to-particle and particle-to-wall contacts were achieved through the combined use of the commercial code and a SolidWorks CAD package. Surface-to-surface radiative heat transfer was modelled in the simulation reflecting real-life application. The numerical results obtained allowed for the determination of parameters such as particle-to-fluid heat transfer coefficient, Nusselt number, Grashof number and Rayleigh number. These parameters were of prime importance when analysing the heat transfer performance of a fixed bed reactor. A comparison of three approaches indicated that the application of the CFD combined with the BUC model gave a better expression of the heat transfer phenomenon in the medium mimicking the heat transfer in the new fuel design / Thesis (PhD)--University of Pretoria, 2016. / Mechanical and Aeronautical Engineering / PhD / Unrestricted

Orientador: João Manoel Losada Moreira / Dissertação (mestrado) - Universidade Federal do ABC, Programa de Pós-Graduação em Energia, 2011

In order to improve the efficiency of dry used nuclear fuel (UNF) storage, reduced reactivity methods are being developed for various reactor types and operating conditions. Sub-criticality must be maintained in the storage configuration and conservative computer simulations are used as the primary basis for loading the storage casks. Methodologies are now being developed to reduce the amount of modeling and computation in order to make conservative assessments of how densely fuel can be packed. The SCALE/TRITON (Standardized Computer Analyses for Licensing Evaluation / Transport Rigor Implemented with Time-dependent Operation for Neutronic Depletion) code system has been used to simulate boiling water reactor (BWR) operating conditions in order to investigate nuclide densities in UNF and how the use of control rod blades affect nuclide densities found in UNF. Rodded and unrodded operating cases for a fuel assembly have been used as bounding cases and are used as reference solutions in a piecewise data approximation methodology (PDA method). A variety of control blade insertion patterns have been used with the PDA method and simulated in TRITON in order to observe trends in nuclide densities with varying control blade use. The PDA method is compared with TRITON simulated data in order to evaluate the validity and accuracy of the PDA method. The PDA method gives very accurate results for fissile nuclides but is insufficient in treating densities as a function of burnup for fission products and fertile nuclides. Predicting nuclide densities from temporally balanced control blade insertion and withdrawal patterns is also a strength of the PDA method. The PDA method, however, is not capable of properly accounting for neutron spectral shifts and the behavior in nuclide densities brought about by the spectral shift or nuclide density saturation. Observing the causes for the shortcomings in the PDA method, a more robust methodology can be developed.

This study aims to improve the modeling of partial SCRAM in the BISON plant model for the Forsmark 2 nuclear reactor after power uprate. Validation of the BISON model against tests performed from March to May in 2013 have shown that this is one of the areas in which there is room for improvement. After partial SCRAM is performed, the model underestimates the reactor power, recirculation flow and steam flow when compared to the measurement data. In BISON the partial SCRAM is modeled using a relative control rod effect vector (ASC vector). The aim is to replace the old values in this vector to improve the model. The new model was shown to give an improved result for the reactor power, recirculation flow and steam flow. The study gives recommendations on how to apply the new model and what values of the relative control rod effect vector that can be used in the future.

No description available.

CANDU (CANada Deuterium Uranium) reactor utilizes Pressure Tube (PT) fuel channel design and heavy water as a coolant. Fuel channel annulus gas acts as an insulator to minimize heat losses from the coolant to the moderator. Since fuel bundles are continuously under high neutron fluxes, annulus gas nuclides undergo different nuclear transformations generating new composition of the gas that might have different physical properties which are undesirable for the annulus system. In addition, gas nuclides become radioactive and lead to an increase of the radioactive material inventory in the reactor and consequently to an increase of radiation levels. Pressure Tube Reactor (PTR) and Pressure Tube Supercritical Water Reactor (PT SCWR) fuel channel models have been developed in Monte Carlo N-Particle (MCNP) code. Neutron fluxes in the fuel channel annulus gas have been obtained by simulating different types of neutron sources in both PTR and PT SCWR fuel channels. Transmutation rates of annulus gases have been calculated for different gases (CO2, N2, Ar and Kr) at different pressures and temperatures in both fuel channels. The variation of the transmutation rates, neutron fluxes and gas densities in the annulus gas have been investigated in PTR and PT SCWR fuel channels at constant pressures and different temperatures. MCNP code along with NIST REFPROP [14] and other software tools have been used to conduct the calculations. / UOIT