Master Theses & Projects (FESNS)

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Analysis and optimal planning of nuclear-renewable hybrid energy systems for ships
(2021-06-01) Adham, Md Ibrahim; Gaber, Hossam
Ocean-going ships are one of the sources of global GHG emissions. Several actions are being taken to reduce the GHG emissions from maritime vessels, and integration of Renewable Energy Sources (RESs) is one of them. Due to some limitations, RESs are not suitable for large ships and often mix with fossil fuel-based generators. Fossil fuel-based generators need to be replaced by emissions-free energy sources to make marine ships free from emissions. Nuclear energy is emissions-free, and small-scale nuclear reactors like Microreactors (MRs) have the potential to replace fossil fuel-based generators. In this study, the technical, economic, and environmental competitiveness of Nuclear-Renewable Hybrid Energy Systems (N-R HES) in marine ships are assessed. The results determine that N-R HES has the lowest NPC compared to the other conventional energy systems. A sensitivity analysis is carried out to see the impact of different system parameters on this study's findings.
The application of experimental microdosimetry to mixed-field neutron-gamma dosimetry
(2012-12-01) Al-Bayati, Saad Najm; Waker, Anthony
Absorbed dose distributions in lineal energy for neutrons and gamma rays were measured by using both a tissue-equivalent walled counter (TEPC) and a graphite-walled low pressure proportional counter (GPC) in the Am-Be neutron source facility at University of Ontario Institute of Technology. A series of measurements were performed with the counters filled with propane-based TE gas (55.1% C3H8, 39.5% CO2 and 5.4% N2) at operating gas pressures corresponding to tissue spheres 2.0 , 4.0 and 8.0 μm in diameter. The results of these measurements indicated satisfactory performance of counters to measure microdosimetric spectra extending down to event-sizes that cover the gamma component of a mixed field. The spectra and the related mean values ̅y F and ̅y D are compared with other similar work but with monoenergetic neutrons of different energy range, the agreement between them is good. An assessment of the performance of different size TEPC has been done. An excellent agreement between their event size spectra was found and the proton edge appears at the same position on the lineal energy scale and differences in microdosimetric parameters ̅ and ̅ is not exceeding 3%, which is in the region of counting statistics. In Am-Be neutron field, the efficiency of the TEPCs was measured to have an average value of 250 counts per μSv or equivalently about 4.17 counts per minutes per μSv/hr. This efficiency is reasonable for dose equivalent measurements but needs a long integration period. The measurements showed that the dose equivalent which depends on the measurement of energy deposition by the secondary charged particles was originated mainly from elastic collisions of the incident neutrons with hydrogen atoms. Moreover the number of events in the sensitive gas is dominated by proton recoils. A non- negligible fraction of the dose equivalent resulted from gamma interactions, alpha and recoil nuclei. The energy deposition patterns in these micro-scale targets are strongly dependent on radiation quality, so differences of linear energy transfer (LET) of the components in a mixed radiation field are significant. Accordingly, in a radiation field with an unknown gamma ray energy spectrum, absorbed dose for neutrons can be obtained by the separation of neutron induced events from gamma events using their distribution in lineal energy. To separate neutron dose from gamma dose a simple lineal energy threshold technique has been used in addition to a more sophisticated methods using γ-fitting and the graphite-walled counter measurements. The results of this study will establish the degree of error introduced by using a lineal energy threshold, which is likely to be used in any hand-held neutron monitor based on TEPCs.
Assessment of FLUENT CFD code as an analysis tool for SCW applications
(2012-08-01) Farah, Amjad; Harvel, Glenn; Pioro, Igor
Chosen as one of six Generation‒IV nuclear-reactor concepts, SuperCritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 ‒ 50% owing to the reactor‟s high pressures and outlet temperatures. The behaviour of supercritical water however, is not well understood and most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations which do not capture the multi-dimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. 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. In this work, the CFD code, FLUENT-12, is used with associated software such as Gambit and NIST REFPROP to predict the Heat Transfer Coefficients at the wall and corresponding wall temperature profiles inside vertical bare tubes with SuperCritical Water (SCW) as the cooling medium. The numerical results are compared with experimental data and 1-D models represented by existing empirical correlations. Analysis of the individual heat-transfer regimes is conducted using an axisymmetric 2-D model of tubes of various lengths and composed of different nodes count along the heated length. Wall temperatures and heat transfer coefficients were analyzed to select the best model for each region (below, at and above the pseudocritical region). To neutralize effects of the rest of the tube on that region, smaller meshes were used were possible. Two turbulent models were used in the process: k-ε and k-ω, with many variations in the sub-model parameters such as viscous heating, thermal effects, and low-Reynolds number correction. Results of the analysis show a fit of ±10% for the wall temperatures using the SST k-ω model in the deteriorated heat transfer regime and less than ±5% for the normal heat transfer regime. The accuracy of the model is higher than any empirical correlation tested in the mentioned regimes, and provides additional information about the multidimensional effects between the bulk-fluid and wall temperatures. Despite the improved prediction capability, the numerical solutions indicate that further work is necessary. Each region has a different numerical model and the CFD code cannot cover the entire range in one comprehensive model. Additionally, some of the trends and transitions predicted are difficult to accept as representation of the true physics of SCW flow conditions. While CFD can be used to develop preliminary design solutions for SCW type reactors, a significant effort in experimental work to measure the actual phenomena is important to make further advancements in CFD based analysis of SCW fluid behaviour.
Characterization of neutron fields around an intense neutron generator
(2016-12-01) Kicka, Leslie; Machrafi, Rachid
Neutron fields in the vicinity of the University of Ontario Institute of Technology neutron facility have been investigated in a series of simulations and experiments. The neutron fluence at several locations around the neutron generator facility has been simulated using MCNPX 2.7E Monte Carlo particle transport program. The P-385 neutron generator is configured to function with a deuterium-deuterium fusion reaction using accelerated charged deuterons colliding with a metal deuteride target. This fusion reaction is characterized by an anisotropic angular and energy distribution in the centre-of-mass and laboratory frames of reference. Three neutron sources were modelled in the simulation with distributions corresponding to different incident deuteron energies of 130 keV, 110 keV, and 90 keV. An idealized isotropic source was likewise simulated for purposes of comparison and determination of the applicability of such an approximation. Along with the performed simulations and to validate the calculation, a series of experiments have been carried out to determine the dose rate measurement at locations adjacent to the generator. The collected data were used to calculate the neutron intensity of the P-385 neutron generator. The measurements were taken using bubble detectors with different sensitivities. Also, the total dose rates corresponding to applied acceleration potentials were estimated at various locations, utilizing a thin target approximation.
Characterization of PVA-Borax-fructose gel for the capture of radioactive wastes generated during nuclear decommissioning activities
(2021-05-01) Sarvendran, Vajran Timothy; Harvel, Glenn
With the long-term objective of using polymer gels for the capture of solid, liquid and gaseous waste generated from decommissioning works, characterization of gel properties as well as the properties of the individual constituents of polymer gels has been completed. Characteristics of constituents were investigated over varied concentrations and environments. Preliminary thermal degradation characterization found no adverse hysteresis behaviour regarding the polymer constituent. Constituent chemistry as well as chemistry that enables crosslinking between constituents was examined, and the theory that most accurately describes the crosslinking of constituents was established. Understanding of chemistry and characterization results were employed to successfully manufacture varied gels, with varied applications. Variation of gels were achieved via adjustments, primarily to polymer concentrations.
Comparative assessment of small modular reactor Passive Safety System design via integration of dynamic methods of analyses
(2019-12-01) Mi, Yi; Tokuhiro, Akira
Integral Pressurized Water Reactor (iPWR) type SMR designs were studied featuring Passive Safety Systems (PSS) in all cases. As many as 11 current SMR designs use PSS to remove decay heat. Variations in PSS designs were studied and compared using evaluation metrics and a proposed weighting method. This resulted in classification of iPWRs designs based on the methodology presented. A prototypic Passive Residual Heat Removal System (PRHRS) was then studied using a scaling analysis to compare the scaling ratio of system parameters, and failure probability relative to existing reference LWR plant data. The impact of single versus two-phase PRHRS designs was also considered. We found that a classical Probabilistic Risk Assessment (PRA) model describing active systems does not consider time evolution nor event ordering that a dynamic PRA approach can accommodate. We thus developed and realized basic coupling between LabVIEW as simulation code and CAFTA as PRA code. Coupling these codes using Python provides real-time simulation that leads to a dynamic simulation result. A representative difference in failure probability using dynamic versus classic PRA revealed that for one, there can be more component demands with different event ordering; thus providing insights into PSS failure probability in the iPWR-type SMR designs. The limitation of the work is essentially in the proprietary details of each SMR design. The value however is in the integrated method of system analysis.
Computational fluid dynamic investigations of flow through an aged CANDU pressure tube
(2021-08-01) Lu, Zheng; Piro, Markus
Single-phase, isothermal computational fluid dynamics simulations of twelve simplified CANDUTM 37M fuel bundles sitting in both as-received and aged pressure tubes were performed with Hydra in this work. The predicted coolant flow behaviour of two cases were compared in order to investigate the impact of pressure tube deformation (sag and diametral expansion) on coolant flow. The boundary conditions and material properties are representative of normal operating conditions in a high-powered channel of the Darlington Nuclear Generating Station. Rigorous mesh and turbulence model sensitivity analyses have been performed in this work. It has been determined that pressure tube deformation has a significant impact on coolant flow behaviour, whereby up to 25% of flow bypasses the fuel. The reduced flow within the subchannels of the fuel bundles impacts the coolability of the fuel.
A Computational Fluid Dynamics based model that predicts wall shear stress in CANDU outlet feeder pipes
(2018-01-01) Wijayaratne, Uditha; Bereznai, George
Wall thinning of carbon steel in CANDU reactor outlet feeder pipes due to Flow Accelerated Corrosion (FAC) is identified as one of the challenges for CANDU reactors since it would force them to shut down due to safety reasons. Several models have been developed over time to predict the corrosion rate (i.e. the rate of wall thinning) of CANDU outlet feeders. These models are developed based on the corrosion chemistry and the mass transfer theories on growth and removal of the protective magnetite layer on the outlet feeder pipe surface. The magnetite layer is acting as a protective layer for the carbon steel feeder pipes by avoiding further corrosion. However, due to the wall shear stress that exerts on the feeder pipe wall, this protective layer is flushed away with the primary heat transport fluid. Wall shear stress is identified as one of the crucial factors behind FAC. Other parameters such as Fe ion concentration, fluid temperature, and pressure would remain within a certain range for a typical CANDU reactor. Still, the distribution of wall shear stress highly depends on the physical arrangement of the outlet feeder pipes. Therefore, wall shear stress would change drastically from one feeder pipe to another resulting in a higher degree of impact on the rate of wall thinning due to FAC. The model developed in this study predicts the maximum wall shear stress on the first bend of a particular feeder pipe considering the fluid Reynolds number, the bend angle and the linear length from the grayloc hub to the first bend. The model is developed using the wall shear stress distribution results generated by Computational Fluid Dynamics (CFD) studies using Siemens NX. The wall shear stress results from the model is then compared against the rate of wall thinning data available for the reactor 01 of the Darlington Nuclear Generating Station as well as some other models available in the literature. iii The model shows a good trend of predicted wall shear stress values against the rate of wall thinning data available. At this stage, the model can be used to identify the feeder pipe with the highest rate of wall thinning due to FAC among a set of given feeder pipes with 2” or 2.5” nominal diameters. This model can be used to identify the optimum feeder pipes for wall thickness measurements during routine maintenance and hence replace the required feeder pipes to avoid any unplanned shut down due to safety reasons.
Conceptual design for a re-entrant type fuel channel for supercritical water-cooled nuclear reactors
(2011-04-01) Samuel, Jeffrey; Harvel, Glenn; Pioro, Igor
Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical Water-Cooled Reactor (SCWR) is to eliminate the annulus gap and use a ceramic insert to insulate the coolant from the moderator. While such a design may work, alternative fuel-channel design concepts are under development to explore the optimum efficiency of SCWRs. One such alternative approach is called the Re-Entrant fuel-channel. The Re-Entrant fuel-channel consists of three tubes, the inner tube (flow tube), pressure tube and an outer tube. The fuel bundles are placed in the inner tube. An annulus is formed between the flow and pressure tubes, through which the primary coolant flows. A ceramic insulator is placed between the pressure tube and the outer tube. The coolant flows through the annulus receiving heat from the inner tube from one end of the channel to another. At the far end, the flow will reverse direction and enter the inner tube, and hence the fuel-string. At the inlet, the temperature is 350°C for a high-pressure coolant (pressure of 25 MPa), which is just below the pseudocritical point. At the outlet, the temperature is about 625ºC at the same pressure (the pressure drop is small and can be neglected). The objective of this work was to design the Re-Entrant channel and to estimate the heat loss to the moderator for the proposed new fuel-channel design. A numerical model was developed and MATLAB was used to calculate the heat loss from the insulated Re-Entrant fuel-channel along with the temperature profiles and the heat transfer coefficients for a given set of flow, pressure, temperature and power boundary conditions. Thermophysical properties were obtained from NIST REFPROP software. With the results from the numerical model, the design of the Re-Entrant fuelchannel was optimized to improve its efficiency
Conceptual design of a thermal hydraulic loop for multiple test geometries at supercritical conditions named supercritical phenomena experimental test apparatus (SPETA)
(2012-04-01) Adenariwo, Adepoju; Harvel, Glenn
The efficiency of nuclear reactors can be improved by increasing the operating pressure of current nuclear reactors. Current CANDU-type nuclear reactors use heavy water as coolant at an outlet pressure of up to 11.5 MPa. Conceptual SuperCritical Water Reactors (SCWRs) will operate at a higher coolant outlet pressure of 25 MPa. Supercritical water technology has been used in advanced coal plants and its application proves promising to be employed in nuclear reactors. To better understand how supercritical water technology can be applied in nuclear power plants, supercritical water loops are used to study the heat transfer phenomena as it applies to CANDU-type reactors. A conceptual design of a loop known as the Supercritical Phenomena Experimental Apparatus (SPETA) has been done. This loop has been designed to fit in a 9 m by 2 m by 2.8 m enclosure that will be installed at the University of Ontario Institute of Technology Energy Research Laboratory. The loop include components to safely start up and shut down various test sections, produce a heat source to the test section, and to remove reject heat. It is expected that loop will be able to investigate the behaviour of supercritical water in various geometries including bare tubes, annulus tubes, and multi-element-type bundles. The experimental geometries are designed to match the fluid properties of Canadian SCWR fuel channel designs so that they are representative of a practical application of supercritical water technology in nuclear plants. This loop will investigate various test section orientations which are the horizontal, vertical, and inclined to investigate buoyancy effects. Frictional pressure drop effects and satisfactory methods of estimating hydraulic resistances in supercritical fluid shall also be estimated with the loop. Operating limits for SPETA have been established to be able to capture the important heat transfer phenomena at supercritical conditions. Heat balance and flow calculations have been done to appropriately size components in the loop. Sensitivity analysis has been done to find the optimum design for the loop.
The conceptualization and parameterization of a gaseous detector rasterizing pinhole gamma camera
(2016-08-01) Price, Terry J.; Machrafi, Rachid
This thesis details the conceptualization and parameterization of a gaseous detector rasterizing pinhole gamma camera. In this thesis, there is a literature review that describes the historical development of gamma imaging, a technical background that aims to give the reader the prerequisite background knowledge, a methodology, and, a result and discussion chapter. The thesis includes studies that determine if the concept of a gaseous detector rasterizing pinhole gamma camera is feasible, mathematical modeling that allowed for the exploration of the idea, software development that automated the mathematical modeling, parametric studies that explored the performance of various sets of design parameters, and, finally an iterative engineering design process that converged at a final set of design parameters. Ultimately, a set of design parameters, from which a prototype may be constructed, were developed.
Decentralized state-space controller design of a large PHWR
(2009-11-01) Khan, Nafisah; Lu, Lixuan
The behaviour of a large nuclear reactor can be described with sufficient accuracy using a nodal model, like the spatial model of a 540 MWe large Pressurized Heavy Water Reactor (PHWR). This model divides the reactor into divisions or nodes to create a spatial model in order to control the xenon induced oscillations that occur in PHWRs. However, being such a large scale system, a 72nd-order model, it makes controller design challenging. Therefore, a reduced order model is much more manageable. A convenient method of model reduction while maintaining the important dynamics characteristics of the process can be done by decoupling. Also, due to the nature of the system, decentralized controllers could serve as a better option because it allows each controller to be localized. This way, any control input to a zone only affects the desired zone and the zones most coupled with, thus not causing a respective change in neutron flux in the other zones. In this thesis, three decentralized controllers were designed using the spatial model of a 540 MWe large PHWR. A decoupling algorithm was designed to divide the system into three partitions containing 20, 27, and 25 states each. Reduced order sub-systems were thus created to produce optimal decentralized controllers. An optimal centralized controller was created to compare both approaches. The decentralized versus centralized controllers’ system responses were analyzed after a reactivity disturbance. A fail-safe study was done to highlight one of the advantages of decentralized controllers.
Design and construction of a one-dimensional particle tracker for measurement of alpha particle stopping power
(2019-08-01) Watt, Sarah; Waker, Anthony
The energy deposited by a charged particle in a medium is non-uniform and peaks near the end of the particle’s path. The energy deposition through the medium is known as stopping power and the shape of this function is called the Bragg curve. Stopping power is typically calculated from first principles rather than measured due to the difficulty of doing so. A one-dimensional particle tracker using gas electron multiplier technology was designed and constructed to directly measure the stopping power at 16 discrete points along the path of alpha particles emitted by 241Am. The use of tissue-equivalent gas allows the results to be compared to those expected within tissue. The results obtained show that the detector concept has merit, although there is room for improvement. In particular, certain voltages and electric field strengths have room for optimization, and more sophisticated readout electronics could be used to reduce experiment run time.
Design of a fault tolerant control scheme based on sliding mode for Canadian supercritical water reactor
(2016-08-01) He, Huan; Lu, Lixuan
Canadian Super Critical Water Reactor (SCWR) is one of the Generation IV reactor types and possible heat source for co-hydrogen production through copper chlorine thermochemical cycle. To maintain the balance between the hydrogen production and electricity production effectively and to utilize the waste heat more efficiently, a well-designed control system for the Canadian SCWR is needed. The SCWR is a nonlinear and strongly coupled multiple-input multiple-output (MIMO) plant. Traditional controller design method, which divides the MIMO system into subsystems and then designs each controller separately, will not obtain satisfactory performance. To accomplish different control objectives of different system variables simultaneously, an integrated multivariable control strategy is needed. Furthermore, as a highly safety-critical system, it is desired that safe and stable performance of the plant can be maintained even in faulty situations. A sliding mode-based Fault Tolerant Control (FTC) scheme, which includes a robust multi-input multi-output controller, a Fault Detection/Isolation (FDI) module and a Fault Accommodation (FA) module, is designed for the overall Canadian SCWR plant in this study. The simulation results indicated that compared to previous control scheme design, better performance was obtained in terms of tracking speed, accuracy, decoupling capacity and control effort in the fault free case. In the faulty case, fault information can be detected and estimated. Acceptable tracking performance was maintained and the variation of steam temperature within design limit was guaranteed.
Detector integration of severe accident management instrumentation for robotic applications at nuclear reactor facilities
(2021-08-01) Nusrat, Omar; Waller, Edward
In the aftermath of a nuclear accident, robots can be used to monitor and assess radiological contamination, preventing harmful exposure to plant personnel. In this work, several detectors were evaluated to be supplemented onto the Husky UGV. Specifically, the RadEye Gamma Survey Meter, the PurpleAir Air-Quality (PA) sensor, and the NaI(Tl) scintillator were examined and their measurement parameters optimized. Optimization was done to satisfy mitigation requirements outlined in regulatory severe accident management guidelines (SAMGs). A software component (Severe Accident Radioactivity Classification; SARC) was developed with the detector components, facilitating detector integration and analysis to aid emergency responders. For the RadEye, 20 seconds was determined to be the optimal collection time; the long term stability and short-term sensitivity of the PA was evaluated; and two spectra measured with the NaI(Tl) were examined. Future work involves further integration of SARC and the addition of advanced capabilities such as infrastructure damage detection.
Determining the effectiveness of nuclear security through computer simulation
(2015-09-01) Chornoboy, Nicholas Jordan; Waller, Edward
There is a growing concern from both national regulators and the International Atomic Energy Agency (IAEA) about the threat posed by attacks against iconic targets such as nuclear power plants. This has led to an increased desire to be able to objectively measure the effectiveness of the physical security of these sites to prevent theft or sabotage of the nuclear and radiological material. Currently verification of physical protection systems is done using subjective expert opinion as well as time consuming and expensive live exercises. A method that allows experts to design and test a facility in the absence of live action exercises using larger sample sizes would be highly desirable. To _ll the niche a synthetic environment model was designed around the force on force simulation program STAGE to allow the full 3-D simulation of a nuclear facility. This allows for simple user modifications to the model, allowing many scenarios to be tested. Many detectors were added to more accurately reflect the types of sensors present at a nuclear facility. Having modeled the facility and the probabilities associated with various events, Monte-Carlo methods were applied to obtain statistics on how effective the guard force was at stopping the adversarial force. This technique can be used to give experts more robust, simple to use tools for the design and verification of physical protection systems.
Developing 1-D heat transfer correlations for supercritical water and carbon dioxide in vertical tubes
(2014-03-01) Gupta, Sahil; Pioro, Igor
Taking into account the expected increase in global energy demands and increasing climate change issues, there is a pressing need to develop new environmentally sustainable energy systems. Nuclear energy will play a major role in being part of the energy mix since it offers a relatively clean, safe and reliable source of electrical energy. However, opportunities for building new generation nuclear systems will depend on their economic and safety attractiveness as well as their flexibility in design to adapt in different countries and situations. Keeping these objectives in mind, a framework for international cooperation was set forth in a charter of Generation IV International Forum (GIF) (GIF Charter, 2002) and six design concepts were selected for further development. To achieve high thermal efficiencies of up to 45 – 50%, the use of SuperCritical Fluids (SCFs) as working fluids in heat transfer cycles is proposed Generation IV designs. An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. In addition to the nuclear power industry applications; SCFs are also expected to play a vital role in a number of other important technologies such as refrigeration systems, and geothermal systems, to name a few. Given the potential for vast number of applications of SCFs in industry, the objective of this work was to gain an understanding on the behavior of SCFs and to develop a fundamental knowledge of the heat-transfer processes and correlations for SC Water and SC CO2 flowing in bare circular tubes. Experimental datasets for SC Water and SC CO2 were compiled and used to obtain a basic 1-D empirical correlation that can predict HTC in bare circular tubes during the transient phases. The accuracy of these correlations was also analyzed using statistical techniques. Limitations and applications for 1-D correlations are discussed as well. The new correlations showed promising results for HTC and Tw calculations for the reference dataset with uncertainty of about ±25% for HTC values and about ±10-15% for the calculated wall temperature.
Developing a framework for thermodynamic measurements of fluoride salt nuclear fuel
(2019-10-01) Hallatt, Daniel Bryan; Piro, Markus
With a single liquid acting simultaneously as a primary coolant and a nuclear fuel, molten salt reactor development has been challenged with understanding the behavior of its complex fluoride solution under various conditions. Experimental thermodynamics can address part of this challenge by providing insights into the system’s behaviour, such as melting points, heat capacities, and solubilities. The current work develops an experimental framework for studying the experimental thermodynamics of fluoride nuclear salt materials. By establishing experimental capabilities and practises, experimental routines are qualified with the goal of providing background on the environment of error of future thermodynamic measurements using differential scanning calorimetry. Infrastructure, standard procedures, a custom crucible, and qualification of a routine for thermodynamic measurements have been developed. Study on particle size reduction, material interactions, and techniques for purification demonstrated a promising environment for thermodynamic measurements.
Development and characterization of a dual neutron and gamma detector
(2011-08-01) Fariad, Abuzar; Machrafi, Rachid
A dual neutron and gamma detection system has been developed for online measurements. The system consists of a single crystal mounted on a photomultiplier tube to detect simultaneously gamma radiation as well as thermal neutrons. A compact data acquisition system has been used for neutron and gamma discrimination. The system has been tested with different gamma energies and with an Am-Be neutron source at the University of Ontario Institute of Technology neutron facility. This thesis presents the characteristics of the developed detector, and experimental data carried out in different experiments in different fields.
Development and characterization of a high resolution portable gamma spectrometer
(2012-04-01) Ali, Muhammad; Machrafi, Rachid
The recent disaster of Fukushima in Japan combined with the high demand to enhance nuclear safety and to minimize personal exposure to radioactive materials has a significant impact on research and development of radiation detection instrumentation. Currently, there is ample effort worldwide in the pursuit of radiation detection to maximize the accuracy and meet international standards in terms of size and specifications to enable radiation protection decision making. Among the requirements is the development of a portable, light-weight gamma-ray isotope identifier to be used by first responders in nuclear accidents as well as for radiation security and identification of illicit material isotopes. From nuclear security perspective, research into advanced screening technologies has become a high priority in all aspects, while for occupational safety, and environmental radiation protection, the regulatory authorities are requiring specific performance of radiation detection and measuring devices. At the applied radiation laboratory of the University of Ontario Institute of Technology the development of a high resolution spectrometer for medium and high energy gamma ray has been conducted. The spectrometer used a newly developed scintillator based on a LaBr3(Ce) crystal. The detector has been modeled using advanced Monte Carlo code (MCNP/X code) for the response function simulation and parameter characterization. The simulation results have been validated by experimental investigations using a wide range of gamma radiation energies. The developed spectrometer has been characterized in terms of resolution and response in different fields. It has also been compared with other crystals such as NaI(TI) and LiI(Eu).

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