Doctoral Dissertations (FESNS)


Recent Submissions

Now showing 1 - 15 of 15
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    A benchmarked dynamic model of xenon behavior in a molten salt reactor
    (2019-12-01) Price, Terry J.; Bereznai, George; Chvala, Ondrej
    Molten salt reactors are a type of nuclear reactor that are being considered for deployment in the fourth generation nuclear power technology. Molten salt reactors use molten a alkali / actinide halide salt melt at temperatures far in excess of temperatures found in a typical pressurized water reactor. This thesis focuses on graphite moderated reactors with fluoride as the halide. The salt melt, called the fuel salt, is circulated between a moderator and a heat exchanger. While within the moderator, the dissolved actinides undergo fission and generate heat. Among products of nuclear fission is gaseous xenon, and in particular the isotope xenon-135 that acts as a neutron absorber. In solid fueled reactors, the xenon is effectively static and trapped within the fuel matrix. In a molten salt reactor, conversely, the fuel matrix is the mobile, circulating fuel salt that transports the xenon along with the rest of the fuel. This thesis focuses on modeling the behavior of xenon in a molten salt reactor. Existing literature in the field is reviewed and compiled. A model of xenon behavior in a molten salt reactor (the Molten Salt Reactor Experiment in particular) has been developed and the model is presented in this thesis. The model is benchmarked against experimental data using best available data, then minimal necessary justifiable adjustment is made to model parameters in order to fit the model to the experimental data. As a result this model is able to fit two transients, something that no xenon model of the molten salt reactor experiment has been able to do previously.
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    New approaches in training for in situ visualization of ionizing radiation measurements
    (2022-08-01) Chaput, Joseph; Waller, Ed
    Surveying an environment for sources of ionizing radiation requires the use of measurement tools and skills developed from training in order to understand and analyze measurements. Recent advances in technology allow for new approaches to be made in the radiation survey methodology generally used with the incorporation of augmented reality (AR) technology to improve real-time awareness in situ and virtual reality (VR) technology to better develop the skill set of the surveyor in realistic virtual environments beforehand. This thesis investigates and develops a novel process to display real-time measured radiation monitoring data in AR to support a radiation surveyor during a search of an environment for hazardous sources of radiation. This AR process is then modified to show how it can be used with virtual radiation sources to allow a radiation surveyor to practice with a digital twin of a radiation field using a virtual source. This process is then modified further and shown how it can be adapted and used to develop VR training scenarios to teach the skill sets needed to assess potential hazards (radiological and non-radiological) and identify sources of radiation. Finally, an approach using reinforcement learning methods is developed and applied to demonstrate a strategy to localize a single radiation source leveraging the real-time measurement data taken in AR.
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    Retrospective assessment of human exposures to low dose ionizing radiation using Electron Paramagnetic Resonance (EPR) dosimetry with tooth enamel
    (2022-04-01) Ghimire, Lekhnath; Waller, Edward
    This study collected the extracted teeth from people of different ages in the Durham region and analyzed them using the X-band CW EPR spectroscopy. The total dose rate from the natural and anthropogenic sources was 1.9721 mSv/year. The anthropogenic dose rate from the various sources was 0.6341 mSv/year, about 47.39% of the natural background dose (1.338 mSv/year) in Durham Region, Ontario. The combined anthropogenic doses from these sources were lower than the local background dose in Durham Region, Ontario, and lower than the regulatory annual effective dose limit of 1 mSv/year in Canada. These data demonstrated that the background doses to the public are lower than the regulatory limit. There is a minimal risk to the public from the anthropogenic doses in Durham Region populations. The dose contribution of the nuclear generating stations is small in Durham Region, Ontario. So, the excess anthropogenic doses could be from diagnostic radiology, nuclear medicines, radiation therapy, and other industrial uses of radiation or radioactive materials, but further study would be needed for conclusions about this region's situation. At the same time, there are more chances of providing deciduous teeth for the low dose (10 - 100 mGy) reconstruction in the actual radiation accidents or chronic exposures. To this end, this study used the dose spiking technique in alanine, where a low dose down to 20 mGy was measured with reasonable precision and accuracy. The same method was used in deciduous teeth for measuring low doses, which was challenging to measure precisely using the conventional EPR methods. The measurement accuracy and reproducibility in the deciduous teeth enamel were significantly higher in that dose range than the conventional methods. Thus, this method can solve the measurement problems associated with low doses and is helpful for retrospective and accident dosimetry. Finally, this study concluded that the total anthropogenic doses in the teeth of Durham Region residents were lower than the regulatory limits. The dose spiking technique can be used to measure low doses in tooth enamel for retrospective and accident dosimetry.
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    Electron paramagnetic resonance for dosimetry in freshwater aquatic environments
    (2020-08-01) Tzivaki, Margarita; Waller, Ed
    This thesis explores the feasibility of detection of radiological contamination in aquatic environments through Electron Paramagnetic Resonance (EPR) spectroscopy. Exposure of freshwater ecosystems is possible as a result of accidents involving facilities of the nuclear fuel cycle. Since this can result in contamination with nuclides such as 90Sr, methods for determining radiation doses in the environment were explored. Specifically for shelled species, the concentration factor for 90Sr was found to have the largest influence on the correct prediction of radiation dose due to this nuclide. Possible doses to zebra mussels due to historic accidents were calculated in excess of 2 Gy in some instances. EPR is a method that is used to detect unpaired electrons that are induced by absorption of ionizing radiation in calcified tissues and as such, it measures lifetime dose to shells. From three organisms investigated, only Dreissenid mussels showed a radiation induced signal under the dose of 20 Gy. A linear relationship of the peak-to-peak height of the line at g = 2:0034 was established and used to improve the quantification of absorbed dose. Shells from different sampling dates were discovered to have different background EPR signals. For the sample group with the lowest background, it was possible to resolve doses as low as 0:2 Gy, thus reducing the value of 2 Gy, previously reported in literature. This provides further validation that EPR dosimetry of shelled species has the potential to contribute to better characterization of absorbed doses in aquatic environments.
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    The influence of surface tension on F2-Bubble Morphology at detachment from carbon electrodes in KF-2HF molten fluoride salt
    (2020-04-01) Seto, Kelvin S. H.; Ikeda, Brian
    Molecular F2 and H2-bubbles were generated electrochemically on amorphous carbon electrodes in KF-2HF molten salt to study the effect of surface tension (ϒ) on the profile of a bubble. F2 is used for the purification and enrichment of fuel in the nuclear industry. Its generation is energy-intensive because of electrical resistances caused by the adhesion of F2-bubbles on the electrode surface. The dimensions and contact angles of generated F2-bubbles were measured and a relationship was developed to study how the profile changes with wetting conditions. F2 was also generated on a large (~4.0 cm2) carbon electrode to observe how the process of bubble formation changes when the size of bubbles at detachment are not restricted by the size of the electrode. The results from these studies revealed inconsistencies with the mixed-phase fluidized layer' model, which could not explain a spherical-to-lenticular transition in bubble shape that was observed. The fluid dynamics of rising bubbles in molten KF-2HF were characterized using the rise velocities (Vt) and aspect ratios (Eb) of F2 and H2-bubbles for a range of bubble sizes. Correlations for Vt, Eb, and drag coefficients (Cd) were evaluated for their accuracy in predicting bubble behavior in this molten salt. A range of -values were obtained by fitting the correlations to the F2 and H2-data sets. A numerical model based on a force balance calculation was developed to estimate the interfacial tension of spherical and lenticular F2-bubbles for different stages of electrode passivation. The calculated ϒGL-value for KF-2HF (0.11 N/m) fit into the trends for surface tension of simple 1:1 and 1:2 molten halide salts. In addition, the GL value was within the range obtained from the correlation fitting. This body of work is the first: 1) study of the effect of wetting on the shape and detachment of lenticular bubbles in a molten salt; 2) comprehensive study of bubble morphology and dynamics in a molten salt; and, 3) estimate of the surface tension of KF-2HF.
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    Developing a thermodynamic model for the U-Pd-Rh-Ru quaternary system for use in the modelling of nuclear fuel
    (2018-12-01) Wang, Lian Cheng; Kaye, Matthew
    Ruthenium, rhodium, and palladium are fission products in nuclear fuels. These elements and their compounds change the properties of fuel pellets. Phase diagrams involving uranium have been constructed experimentally to study fission product behaviour, specifically diagrams containing the very stable UMe3 (where Me = Ru, Rh, or Pd) compounds. Discrepancies such as enthalpies of formation of UxMey compounds exist in both experimental binary phase diagram constructions and thermodynamic property determinations. To model the behaviour of fission products in irradiated nuclear fuels, codes (e.g., BISON or RNFTT (RMC Nuclear Fuel Thermochemical Treatment) have been developed. For quantitative studies, existing experimental data are insufficient for such tasks because of difficulties determining data in ranges of composition and temperature. Experimental binary phase diagrams provide phase equilibrium information, yet if not thermodynamically evaluated, the data will be limited in application. Because industrial processes usually involve multicomponent systems that vary in wide ranges of composition and temperature. For some elements with potential catalytic functions (e.g., Pd), the U-Pd phase diagrams presented in the literature were inconsistent so it was a challenge to choose which experimental data should be used for a thermodynamic evaluation. Post irradiation examinations showed that the composition of irradiated nuclear fuels are complicated. For such complex systems, experimental determination of a full set of data is practically impossible. Nevertheless, the possibility of constructing such complex systems by means of thermodynamic evaluation exists. In this work, thermodynamic evaluations of the URu, U-Rh, and U-Pd binary phase diagrams were assessed or re-assessed (e.g., the U-Ru system). In combination with three binary systems previously assessed, a self-consistent quaternary system (U-Pd-Rh-Ru) was constructed. An alternative strategy in optimizing the Gibbs energy functions of various phases, capable of identifying experimental fallacies in hand drawn U-Rh and U-Pd phase diagrams, was proposed. With this quaternary model, two existing ternary experimental phase diagrams were critically evaluated. Results show that without thermodynamic evaluations some experimental data were wrongly interpreted. The establishment of the quaternary model enriches thermodynamic databases and will potentially improve the performance of the RNFTT treatment and codes such as BISON.
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    Methods for optimization of neutron detector performance in nuclear power plants
    (2018-08-01) Nasimi, Elnara; Gaber, Hossam
    Safety of nuclear power plants (NPPs) requires shutdown systems to mitigate design-basis accidents. Shutdown is based on measurements from in-core flux detectors. Thus, this thesis proposes a methodology for managing neutron-detector aging. Detector performance improvement for prompt fraction along with genetic algorithm optimization, risk estimation, and the fault semantic network (FSN) are studied. Detector models have been developed using circuit models to represent detector behavior in aged conditions. For cases with preprocessed measured data, initial conditions are set to the initial sample value, and an inverse of the transform is applied to reconstruct the signal at the detector. The model’s response is compared to the plant data to validate output. Proposed gains have been implemented in the model with the old gain values to compare with FUELPIN code predictions. A genetic algorithm has been selected to solve optimization and search problems where the Dynamic Signal Compensator (DSC) function is the objective function and the detector function is the fitness function. Two new sets of gain constants have been used to solve the power ramp-up case, allowing reduction of maximum margin-to-trip (MTT) loss to 0.8%. The risk model has been established in the FSN using historical data, and the risk factor for all channels has been identified. The initial channel table and risk prioritization map are updated as information becomes available. This is demonstrated in three case studies where assumed component failure rates were used in a qualitative/quantitative manner for risk estimation. The results suggest the simulated ramp rate with the optimized DSC gain constants is conservative compared to actual fueling and signal oscillation rates. Further gains in detector prompt fraction were obtained by optimization of the amplifier gain settings through genetic algorithm with no adverse impact on operating margins.
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    Investigation of the bubble detector response to high LET space radiation
    (2018-08-01) Miller, Alexander L.; Machrafi, Rachid
    The radiation environment aboard spacecraft is a complex mixture of neutrons, photons, protons, heavy ions and other particles. A special type of superheated droplet detectors referred to as space bubble detectors (SBD) have been used to evaluate the equivalent dose due to neutrons in various space missions aboard the International Space Station. Protons and other heavy charged particles are a significant component of the high LET radiation field and also contribute to the SBD measurements. The calibration of the bubble detectors is established using a known Americium Beryllium(AmBe) neutron field. However, the space neutron field is considerably different from the AmBe field. Current models assume that bubbles are formed as a result of radiation interactions above a certain minimum LET threshold and experiments have shown that the LET threshold may be different for different ions.In order to interpret the bubble detector measurements in space radiation fields, a systematic investigation of the response of bubble detectors to high LET radiation encountered inspace has been performed. A series of experiments have been conducted with different high LET radiation including protons and energetic heavy ions using different facilities at the National Institute of Radiological Science in Chiba, Japan, and the ProCure Proton Therapy Center in Oklahoma, USA. High energy neutron experiments were conducted at the Los Alamos Neutron Science Center. A correction factor of 1.8 ± 0.2 has been determined to correlate the AmBe calibrated sensitivity to neutron equivalent dose measurements aboard the ISS. The LET threshold required to form a bubble in SBD was found to depend on the charge Z of the ion. An analytical model to evaluate the SBD response to high LET radiation aboard the ISS has been developed and compared to measurements.
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    Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs)
    (2017-08-01) Peiman, Wargha; Pioro, Igor; Gabriel, Kamiel
    A group of countries has initiated an international collaboration to develop a next generation (i.e., Generation IV) of nuclear reactors. Chosen as one of the six Generation‒IV nuclear-reactor concepts, the SCWRs are expected to have high thermal efficiencies within the range of 40 ‒ 50% owing to reactor’s high outlet temperatures. The Canadian pressure-tube-type SCWR is featured with 3-batch refueling, 336 vertical fuel channels, a porous ceramic insulator inside the pressure tube, and stainless-steel cladding. The reactor operates at a pressure of 25 MPa with the coolant temperature rising from 350 to 625°C. Consequently, sheath and fuel centerline temperatures are significantly higher in SCWRs compared to those of the current water-cooled nuclear reactors. The main objective of this thesis is to conduct a study on specifics of the thermalhydraulics and neutronics of a pressure-tube SCWR based on an understanding of the supercritical water phenomena and their impacts on reactor design and operation. This thesis investigates the impact of several thermalhydraulic modeling parameters on fuel and cladding temperatures of a pressure-tube SCWR. The investigated thermalhydraulic modeling parameters are: 1) variable heat transfer coefficient, which is affected by thermophysical properties of supercritical water, axial heat flux, and three heat-transfer regimes: normal, improved and deteriorated; 2) thermophysical properties, which are affected by the bulk-fluid-temperature profile along the heated length and pressure drop along the fuel channel; 3) variable axial and radial heat-flux profiles of a fuel assembly (bundle string), which are affected by the neutron flux; 4) radial non-uniform heat generation inside the fuel; 5) axial and radial variable thermal conductivity of a fuel; 6) contact thermal resistance between the fuel and cladding; 7) heat loss from the coolant to the moderator, which is affected by the thermal conductivity of a ceramic insert; and 8) pressure drop of the coolant along the fuel channel. The main neutronic aspects, which have been incorporated in the neutronic model, include 1) variable coolant density along the heated length of the fuel channel, which affects neutronic properties of a lattice and, hence, the neutron flux and 2) number of energy groups, which affects the calculated channel powers.
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    CFD determination of fluid and geometry related localized heat transfer phenomena for supercritical water flow
    (2017-04-01) Farah, Amjad; Harvel, Glenn; Pioro, Igor
    The proposed concept of Supercritical Water-cooled Reactor (SCWR) as part of the Generation IV International Forum aims to improve the thermal efficiency over current power plants by utilizing cooling water at pressures and temperature above the critical point. At supercritical conditions, however, the properties of the fluid can vary rapidly in response to changes in temperature and pressure, and without a phase change. One example is the specific heat, which exhibits a sharp peak at a point defined as the pseudocritical temperature. Computational Fluid Dynamics (CFD) is a numerical approach to model fluids in multidimensional space using the Navier-Stokes equations and databases of fluid properties to arrive at a full simulation of a fluid dynamics and heat transfer system. Turbulence models employed in CFD are a set of equations that determine the turbulence transport terms in the mean flow equations. They are based on hypotheses about the process of turbulence, and as such require empirical input in the form of constants or functions, in order to achieve closure. This work is conducted to further develop an understanding of supercritical water (SCW) flow by analyzing the flow- and geometry-dependent localized phenomena under supercritical conditions using CFD turbulence models. The numerical study employed the Realizable k-ε and the SST k-ω turbulence models. The created meshes are three dimensional to capture the multi-dimensional effects of SCW heat transfer phenomena. In the first part of the study, the turbulent Pr number effect on SCW heat transfer characteristics is determined by analyzing changes in fluid properties such as temperature profiles, turbulence intensity, and velocity in response to varying the turbulent Pr values in the CFD models. This investigation has shown the energy turbulent Pr to have the most effect on improving SCW heat transfer simulation results under the deteriorated heat transfer regime, by affecting the turbulence production in the fluid due to buoyancy forces. Buoyancy forces were also studied in downward flow under the same conditions and were shown to reduce the deterioration in heat transfer observed in upward flow. The second part involved an investigation of fluid property effects in complex geometries to determine important flow parameters that capture localized flow phenomena effects. Two geometries are considered: an annular channel with helical fins, and a tube with a sudden area change. The helicity of the first geometry did not appear to induce additional turbulence in the flow, compared to bare geometries. On the other hand, the sudden area change introduced large levels of turbulence, and while it dissipated quickly, it did show an enhancement in the heat transfer and lowered the outlet wall temperatures. These results can be used as a design input for SCWR fuel geometry design. As a result, this study contributes to the understanding of the SCW heat transfer fundamentals under normal and deteriorated regimes in bare and complex geometries, and identifies the areas of improvement in the related experimental work. Significant experimental work is needed to verify the findings
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    Dynamic safety assessment of FPGA-based safety critical systems with applications in nuclear power generation
    (2016-12-01) McNelles, Phillip; Lu, Lixuan
    Field Programmable Gate Arrays (FPGAS) are a type on integrated circuit that is configured by the end user to perform desired digital logic functions. FPGAs do not run any software or operating system, as the logic functions are configured as a hardware implementation on the FPGA chip. Documentation from the International Atomic Energy Agency (IAEA) states that FPGA implementations of I&C systems in Nuclear Power Plants (NPPs) is expected to increase significantly in the future. One issue facing FPGAs in the nuclear field is a lack of technical standards and design/review documentation. Therefore, the research program undertaken during this thesis considered the application of a new safety analysis methodology for the modelling and analysis of FPGA-based systems. The methodology chosen is a modern, dynamic (time-dependant) methodology known as the Dynamic Flowgraph Methodology (DFM), which is intended to be applied to digital I&C systems. Initially, a Failure Modes and Effects Analysis (FMEA) was performed to ascertain the potential failure modes that could affect FPGA-based systems, and that FMEA data was used to create and FPGA failure modes taxonomy. Using that FMEA data to provide information for fault injection, DFM was applied to analyze several FPGA-based test systems, and the results of the DFM analyses were compared and contrasted with results from Fault Tree Analysis (FTA), to determine the potential advantages and disadvantages of DFM. It was seen that DFM had several advantages when modelling clock delays, oscillating clock signals, and Multiple-Valued Logic, however for large systems DFM continues to experience the “state explosion” problem, limiting its effectiveness to small-medium sized systems. Potential avenues of future work are also presented.
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    An evaluation of the net breeding capability of heterogeneously-fuelled pressure-tube heavy water reactors with the thorium fuel cycle
    (2015-12-01) Golesorkhi, Sourena; Kaye, Matthew Haigh
    The thorium fuel cycle is a promising future option for an alternative to uranium. The pressure tube heavy water reactor (PT-HWR) has for decades been considered capable of achieving net breeding on the thorium fuel cycle, but this capability has not been conclusively demonstrated. The goal of this work was to perform reactor physics modelling of the 380-channel, 700 MWe PT-HWR, attempting to achieve a self-sustaining equilibrium thorium cycle with specific focus on operational viability. The DRAGON neutron transport code was used to model ThO₂/²³³UO₂ fuel lattice cells. Sensitivity analyses were performed for the fissile nuclide content, specific power, and average fuel temperature. Thorium-based fuels are found to have a strongly negative power coefficient of reactivity, leading to criticality concerns in the event of a prolonged shutdown. Several fuel bundle concepts were modelled in order to determine the most favourable assembly for breeding. The results showed no significant benefit to other concepts, therefore the standard 37-element bundle was selected due to its wealth of operational experience and well-known operating margins. Homogeneous fuelling was found to be impractical for breeding, and heterogeneous fuel bundles were found to not offer significant improvement while increasing complexity. As a result, heterogeneous core configurations using homogeneous bundles were investigated in full core calculations. The DONJON core physics code was used in conjunction with homogenized cross-sections calculated by DRAGON to model the complete reactor, including reactivity devices and supporting structures. Seven fuelling configurations were simulated and iteratively improved through an empirical process. Ultimately, none of the studied configurations could achieve net breeding. The most favourable configuration was found to be a heterogeneous seed and blanket core where the blanket (composed of 1.4 at% ²³³U) was placed in the central region and the seed (composed of 1.6 at% ²³³U) was placed in the periphery. Two variants of this configuration (differing on the refuelling scheme used in the blanket) were further investigated with instantaneous power simulations with 10 full power days of refuelling. Both variants were found to abide by the existing license limits on maximum bundle and channel power. The variant using 4-bundle shift in both blanket and seed could tolerate an increase in reactor power to 110% FP while maintaining a comfortable margin to safety. A number of postulated misfuelling events were simulated for both variants, and the responses were found to be controllable by the existing reactivity systems. It is noted that there are significant sources of error in the results of this work. Advances in computational methods as well as better nuclear data for the thorium fuel cycle are required for more accurate predictions. Overall, while the PT-HWR has great potential as a very high converting reactor, based on the results of this work, the existing design cannot operate as a net breeder. From an economic perspective, a gain in power output may be more beneficial than an entirely closed fuel cycle.
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    Specifics of forced-convective heat transfer to supercritical CO2 flowing upward in vertical bare tubes
    (2015-04-01) Saltanov, Eugene; Pioro, Igor; Harvel, Glenn
    Heat transfer in the forced convection regime of fluids at supercritical conditions has been studied extensively for the past 60 years. The dominant approach to summarize the experimental results was by proposing empirical correlations for the data within the investigated range of parameters. It was soon realized by researchers worldwide that heat transfer coefficients become non-linear functions of wall and bulk-fluid temperatures at certain combinations of experimental parameters within the region of the peak of specific heat at supercritical pressures. Thus, it has become a standard approach to remove nonlinear experimental heat transfer coefficient values treating them as a sign of a deteriorated (as opposed to normal) heat transfer regime. There were recent attempts to address this shortcoming and extend the applicability of conventional empirical correlations to the deteriorated heat transfer regime. However, these attempts were not satisfactory. In this thesis, a new methodology has been developed that allows the use conventional empirical correlations without distinguishing entrance effects or deteriorated heat transfer regime. The methodology is based on binning experimental data according to the parameter X = (h_b - h_pc) / (q/G) and then combining correlations based on wall and bulk-fluid temperature on each bin to minimize RMS and maximal overprediction of heat transfer coefficients within each of the bins. Using this methodology, 95% of normal heat transfer data were predicted with a spread of ±19%, which is 1.74 times narrower compared to the prediction by the empirical correlations developed based on the conventional methodology and on the same data; while all the data (2786 points, including entrance effects and deteriorated heat transfer) were predicted with a spread of ±20% (based on 2σ-level). The data correlated based on the new methodology where obtained within the following range of experimental parameters: P = 7.58 – 8.91 MPa, Tb = 20 – 142 ˚C, Tw = 32 – 231 ˚C, G = 885 – 3048 kg/m2s, q = 26 – 616 kW/m2K, D = 8.1 mm. The experimental data were obtained based on a series of tests on supercritical CO2 flowing upwards in a bare tube at the MR-1 loop (located in Chalk River) of the former Atomic Energy of Canada Limited (AECL). Normal, deteriorated, and improved heat transfer regimes were covered in the experiments.
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    Characterization of eutectic In-Bi-Sn alloy (Field’s metal) for use in single and two-phase liquid metal flow in natural circulation systems
    (2015-04-01) Lipchitz, Adam; Harvel, Glenn
    An In-Bi-Sn eutectic alloy was characterized for the purpose of liquid metal natural circulation experiments. The alloy was chosen to reduce the power requirement and enhance the safety of these types of experiments. The initial characterization included the development of a method of fabrication, an investigation of the chemical compatibility with air, water, borosilicate glass, and stainless steel. An investigation to experimentally determine the thermo-physical properties (viscosity, density, specific heat capacity, and thermal conductivity) of the liquid metal was performed. A numerical model of the natural circulation loop was developed and compared to experimental results of a liquid metal natural circulation experiment.
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    A new approach to neutron spectrometry with multi-element scintillators
    (2014-02-01) Khan, Nafisah; Machrafi, Rachid
    The combined effects of the nuclear industrial renaissance, the events of 9/11, and the Fukushima disaster have had a significant impact on the research and development of radiation detection instrumentation. Notably, there is ample worldwide scientific research effort into a better understanding of the material properties and nuclear interactions with a view to support the improvement of radiation detection and measurement. These improvements are spurred by the heightened security requirements that entail the monitoring of contraband material including explosives in transit, and the need to enhance occupational safety as well as environmental radiation protection in respect to nuclear power generation. Moreover, the regulatory authorities require routinely employed radiation detection and measuring devices in nuclear installations to meet the revised standard specifications in terms of their design and performance. Currently, there is no neutron dosimeter/spectrometer that can meet the requirements in terms of size and performance. In particular, the requirements for high sensitivity, spectroscopic features, and determination of operational quantities to enable radiation protection decision making have resulted in a closer examination of the basic physics of the radiation interactions in detector materials. Neutron dosimetry is regarded as the last frontier in radiation protection. Due to the large span of neutron energy and the strong energy dependence of the dose to fluence coefficient, neutron dosimetry requires the knowledge of the neutron spectra for any accurate neutron dose quantification. As a result, spectrometry is a precursor to determine dosimetry quantities and spectrometers are therefore vital to determine and characterize radiation fields present to individuals as they provide information about the radiation intensity and energy spectra. However, current spectrometers have many drawbacks and limitations in different aspects. From one side, fast spectrometry currently uses the scattering process on hydrogen rich materials and uses complicated unfolding techniques to extract the energy spectra. From another side, neutron fields are inherently mixed with a gamma component and therefore, it is paramount to distinguish each component since their contribution to the dose equivalent is weighted differently. More specifically, the challenge becomes more profound with neutrons in the energy range between 10 keV and few MeV. These challenges are mainly due to: • The drastic change in the dose-to-fluence conversion coefficient that increases by a factor of 40; • The low sensitivity of the neutron sensors used in neutron spectroscopy (low cross section); • The poor resolution of the detectors, which makes accurate neutron spectrometry difficult to achieve. However, by exploiting new developments and high sensitivity scintillators, in this thesis, a new approach has been adopted using different nuclear reaction processes with different contents of scintillating material. More specifically, two nuclear reactions, i.e. (n,α) and (n,p), on two different elements have been used to carry out neutron spectrometry. In addition, this thesis aims to investigate the spectrometric properties of scintillating materials as a first step to establish a platform for developing a neutron spectrometer/dosimeter. In terms of methodology, the thesis has taken an empirical approach in studying potential sensors that can be used for neutron spectrometry. Four scintillators have been explored and studied. Each scintillator corresponds to a particular energy region. The first part of the thesis consists of extensive Monte Carlo calculations to optimize the sensor’s isotope contents, while the second part consists of conducting a series of experiments using three main facilities, namely an AmBe source of 120 mCi, a neutron generator of 2.5 MeV neutrons at the University of Ontario Institute of Technology, a KN Van De Graaff accelerator at McMaster University, and gamma ray sources with different energies. All sensors have been used in conjunction with a miniature data acquisition system that consists of a multi-channel analyzer, mounted on a photomultiplier, and controlled by software to operate, control and analyze the output data. The time characteristics of the output pulse, such as integration time and rising time, have been optimized for each sensor. The thesis presents a thorough literature review, a comprehensive methodology of the study, a description of the used facilities and the results of the simulation data with four different sensors along with the experimental work carried out at the aforementioned facilities. The response functions of each scintillator to a given radiation type and energy has been analyzed and discussed. Furthermore, the sensor’s potential for use in neutron spectrometry/dosimetry has been assessed for future work.