At head of title: Energy Laboratory and Dept. of Nuclear Engineering. / Sponsored by Boston Edison Company and others under MIT Energy Laboratory Electric Utility Program.

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>

No description available.

Thesis (doctoral)--Lappeenranta University of Technology, 2005. / Includes bibliographical references. Also available on the World Wide Web. Myös verkkojulkaisuna.

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

The Multi-Application Small Light Water Reactor (MASLWR) is a small natural circulation pressurized light water reactor design that was developed by Oregon State University (OSU) and Idaho National Engineering and Environmental Laboratory (INEEL) under the Nuclear Energy Research Initiative (NERI) program to address the growing demand for energy and electricity. The MASLWR design is geared toward providing electricity to small communities in remote locations in developing countries where constructions of large nuclear power plants are not economical. The MASLWR reactor is designed to operate for five years without refueling and with fuel enrichment up to 8 %. In 2003, an experimental thermal hydraulic research facility also known as the OSU MASLWR Test Facility was constructed at Oregon State University to examined the performance of new reactor design and natural circulation reactor design concepts. This thesis is focused on the thermal hydraulics analysis and fuel performance analysis of the MASLWR prototypical cores with fuel enrichment of 4.25 % and 8 %. The goals of the thermal hydraulic analyses were to calculate the departure nucleate boiling ratio (DNBR) values, coolant temperature, cladding temperature and fuel temperature profiles in the hot channel of the reactor cores. The thermal hydraulic analysis was performed for steady state operation of the MASLWR prototypical cores. VIPRE Version 01 is the code used for all the computational modeling of the prototypical cores during thermal hydraulic analysis. The hot channel and hot rod results are compared with thermal design limits to determine the feasibility of the prototypical cores. The second level of analysis was performed with a fuel performance code FRAPCON for the limiting MASLWR fuel rods identified by the neutronic and thermal hydraulic analyses. The goals of the fuel performance analyses were to calculate the oxide thickness on the cladding and fission gas release (FGR). The oxide thickness results are compared with the acceptable design limits for standard fuel rods. The results in this research can be helpful for future core designs of small light water reactors with natural circulation. / Graduation date: 2012

The United States' fleet of Light Water Reactors (LWRs) produces a large amount of spent fuel each year; all of which is presently intended to be stored in a fuel repository for disposal. As these LWRs continue to operate and more are built to match the increasing demand for electricity, the required capacity for these repositories grows. Georgia Tech's Subcritical Advanced Burner Reactor (SABR) has been designed to reduce the capacity requirements for these repositories and thereby help close the back end of the nuclear fuel cycle by burning the long-lived transuranics in spent nuclear fuel. SABR's design is based heavily off of the Integral Fast Reactor (IFR). It is important to understand whether the SABR design retains the passive safety characteristics of the IFR. A full safety analysis of SABR's transient response to various possible accident scenarios needs to be performed to determine this. However, before this safety analysis can be performed, it is imperative to model all components of the reactivity feedback mechanism in SABR. The purpose of this work is to develop a calculational model for the fuel bowing reactivity coefficients that can be used in SABR's future safety analysis. This thesis discusses background on fuel bowing and other reactivity coefficients, the history of the IFR, the design of SABR, describes the method that was developed for calculating fuel bowing reactivity coefficients and its validation, and presents an example of a fuel bowing reactivity calculation for SABR.

The subject of this thesis is the development of a nodal discretization of the low-order quasi-diffusion (QDLO) equations for global reactor core calculations. The advantage of quasi-diffusion (QD) is that it is able to capture transport effects at the surface between unlike fuel assemblies better than the diffusion approximation. We discretize QDLO equations with the advanced nodal methodology described by Palmtag (Pal 1997) for diffusion. The fast and thermal neutron fluxes are presented as 2-D, non-separable expansions of polynomial and hyperbolic functions. The fast flux expansion consists of polynomial functions, while the thermal flux is expanded in a combination of polynomial and hyperbolic functions. The advantage of using hyperbolic functions in the thermal flux expansion lies in the accuracy with which hyperbolic functions can represent the large gradients at the interface between unlike fuel assemblies. The hyperbolic expansion functions proposed in (Pal 1997) are the analytic solutions of the zero-source diffusion equation for the thermal flux. The specific form of the QDLO equations requires the derivation of new hyperbolic basis functions which are different from those proposed for the diffusion equation. We have developed a discretization of the QDLO equations with node-averaged cross-sections and Eddington tensor components, solving the 2-D equations using the weighted residual method (Ame 1992). These node-averaged data are assumed known from single assembly transport calculations. We wrote a code in "Mathematica" that solves k-eigenvalue problems and calculates neutron fluxes in 2-D Cartesian coordinates. Numerical test problems show that the model proposed here can reproduce the results of both the simple diffusion problems presented in (Pal 1997) and those with analytic solutions. While the QDLO calculations performed on one-node, zero-current, boundary condition diffusion problems and two-node, zero-current boundary condition problems with UO₂-UO₂ assemblies are in excellent agreement with the benchmark and analytic solutions, UO₂-MOX configurations show more important discrepancies that are due to the single-assembly homogenized cross-sections used in the calculations. The results of the multiple-node problems show similar discrepancies in power distribution with the results reported in (Pal 1997). Multiple-node k-eigenvalue problems exhibit larger discrepancies, but these can be diminished by using adjusted diffusion coefficients (Pal 1997). The results of several "transport" problems demonstrate the influence of Eddington functionals on homogenized flux, power distribution, and multiplication factor k. / Graduation date: 2003

Oregon State University has hosted an International Atomic Energy Agency (IAEA) International Collaborative Standard Problem (ICSP) through testing conducted on the Multi-Application Small Light Water (MASLWR) facility. The MASLWR facility features a full-time natural circulation loop in the primary vessel and a unique pressure suppression device for accident scenarios. Automatic depressurization system (ADS) lines connect the primary vessel to a high pressure containment (HPC) which dissipates steam heat through a heat transfer plate thermally connected to another vessel with a large cool water inventory. This feature drew the interest of the IAEA and an ICSP was developed where a loss of feedwater to the steam generators prompted a depressurization of the primary vessel via a blowdown through the ADS lines. The purpose of the ICSP is to evaluate the applicability of thermal-hydraulic computer codes to unique experiments usually outside of the validation matrix of the code itself. RELAP5-3D 2:4:2 was chosen to model the ICSP. RELAP5-3D is a best-estimate code designed to simulate transient fluid and thermal behavior in light water reactors. Modeling was conducted in RELAP5-3D to identify the strengths and weaknesses of the code in predicting the experimental trends of the IAEA ICSP. This extended to nodalization sensitivity studies, an investigation of built-in models and heat transfer boundary conditions. Besides a qualitative analysis, a quantitative analysis method was also performed. / Graduation date: 2013