Industrial dynamical, Networked Control Systems (NCSs) are controlled over a communication network. We study a continuous-time CANada Deuterium Uranium-Super Critical Water Reactor (CANDU-SCWR) hydrogen plant and a discrete-time controller, sensor and actuator block, that are connected via a communication network, such as e.g. controller area network (CAN), Ethernet or wireless networks. Issues associated with NCSs are time-varying delays, timevarying sampling intervals and loss of data due to packet drop outs. Delays are also associated with software chosen, control system architecture and computation load. CANDU-SCWR hydrogen co-generation model reliability can be analyzed by dynamic flow graph methodology. We have analyzed the CANDU-SCWR feed water integration with the oxygen unit of copper chloride cycle and also conducted an analytical review of the current networked control system delays. / UOIT

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

The Oak Ridge Isotope Generation and Depletion – Automatic Rapid Proccessing (ORIGEN-ARP) deterministic code has been extensively utilized for determining nuclide concentrations at various specific burnup values for a variety of nuclear reactor designs. Given nuclide concentrations or ratios, such calculations can be used in nuclear forensics and nuclear non-proliferation applications to reverse-calculate the type of reactor and specific burnup of the fuel from which the nuclides originated. Recently, Los Alamos National Laboratory has released a version of its probabilistic radiation transport code, MCNPX 2.6.0, which incorporates a fuel burnup feature which can also determine, via the probabilistic Monte Carlo method, nuclide concentrations as a function of fuel burnup. This dissertation compares the concentrations of 46 nuclides significant to nuclear forensics analyses for different reactor types using results from the ORIGEN-ARP and the MCNPX 2.6.0 codes. Three reactor types were chosen: the Westinghouse 17x17 Pressurized Water Reactor (PWR), the GE 8x8-4 Boiling Water Reactor (BWR), and the Canadian Deuterium Uranium, CANDU-37, reactor. Additionally, a sensitivity study of the different reactor parameters within the MCNPX Westinghouse 17x17 PWR model was performed. This study analyzed the different nuclide concentrations resulting from minor perturbations of the following parameters: assembly rod pitch, initial moderator boron concentration, fuel pin cladding thickness, moderator density, and fuel temperature. / text

A Barkhausen noise (BN) testing system was developed for the non-destructive evaluation (NDE) of residual stresses in CANDU reactor feeder pipes. The system consists of a four-channel arbitrary waveform flux control system (FCS), and the spring-loaded tetrapole prototype (SL4P) BN probe. The combination of the FCS and SL4P was shown to provide repeatable BN measurements on feeder pipe samples, with variations in air gaps between the SL4P poles and the sample from 0.43 mm to 1.29 mm, and typical pickup coil coupling uncertainties for the total BN energy from ±2% to ±7%. Precision for elastic strain estimation in feeder pipes was found to be between ±7 MPa and ±9 MPa in tension, depending on the excitation field configuration, and negligible in compression. Modelling of the BN penetration depth as a function of the excitation field was used to estimate the BN penetration depth between 5 μm at 300 kHz to a maximum of 500 μm at 3 kHz. The modelling, engineering, and procedures developed for this BN testing system provide an improved basis for the future advancement of BN testing, and ferromagnetic NDE in general. / Thesis (Ph.D, Physics, Engineering Physics and Astronomy) -- Queen's University, 2009-07-22 15:34:28.967

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One of two project reports: The other part is designated Part B: On-Campus Project / <p> Computer studies are made of a CANDU-PHW reactor core containing an annulus of enriched uranium around a central zone of natural uranium. For hybrid cores of this type with a maximum radial form factor, the uranium requirements, fuel costs, stability, and power peaking upon refuelling are investigated. It is found that these hybrid cores offer potential savings of 10% to 20% in fuel costs and uranium utilization compared to the present CANDU-PHW core, and are only slightly less stable. However, power peaking upon refuelling is a problem with these cores.</p> / Thesis / Master of Engineering (ME)

This report describes computer studies which were done to determine the effect on burnup of modifying the heavy water reflector in a 600 MWe CANDU-PHW reactor. It is shown that the burnup penalty increases rapidly as the reflector thickness is reduced. The burnup penalty is significantly lower for mixed reflectors in which some of the heavy water in the outer region of the reflector is replaced by graphite, an organic liquid, or light water, while maintaining the original reflector thickness. / Thesis / Master of Engineering (MEngr)

With the complex composition of the radioisotopes and waste materials, the characterization of the volumetric low-level wastes from CANDU plants is challenging. This study presents a technique to localize and quantify the contaminations presented in the CANDU waste containers. MCNP-based models are developed for an N-type coaxial HPGe detector and a LaBr3 detector to simulate the photon peak information. The simulated efficiency and the experimental count rates are combined to estimate the activity of unknown waste samples. During the spectrum collection of a 4L Marinelli beaker source and 1-quart waste samples, the MCNP algorithm showed better accuracy in activity estimation than the Mirion ISOCS/LabSOCS software. With further development, this method has the potential to outperform the popular commercial software in estimating activity for volume sources with complex geometry and uneven distribution. The multi-detector array models with hotspot designs are also studied in this work to provide real-time information about the location and activity of the contamination inside the 2.2 m3 industrial low-level waste containers. The on-site measurements show promising results as the position of the contamination was able to be located within a volume of 61×40×34 cm. Overall, this technique has good potential to be utilized in the nuclear industry for large-volume low-level waste analysis. / Thesis / Master of Science (MSc)

<p>A study was conducted to scope the relative merits of various thorium fuel cycles in CANDU reactors. It was determined that, due to the very large reprocessing demands of the self-sustaining equilibrium thorium fuel cycle, an additional fissile driver fuel is required for a practical thorium fuel cycle. The driver fuels considered were PWR- and CANDU- derived plutonium, PWR-derived MOX fuel, and low-enriched uranium. The addition a RU-fuelled CANDU reactor with possible americium, curium, and lanthanides recycling was also considered. The fuel cycles were evaluated for natural uranium consumption and reprocessing demands as well as spent fuel characteristics such as: thermal and gamma power, radioactivity, and ingestion and inhalation hazards.</p> <p>The two-dimensional multigroup code, WIMS-AECL, was used to calculate the burnup and some controllability properties of the CANDU reactors. ORIGEN, a depletion and decay module, was used to evaluate the spent fuel characteristics and the systems code, DESAE, was used to simulate the introduction of the thorium fuel cycle to a growing global reactor park.</p> <p>It was determined that ²³³U production in the thorium fuels is optimized at lower exit burnups. Therefore, less external fissile driver material is required for the operation of the thorium reactors and natural uranium savings of the overall fuel cycle are increased. Furthermore, it was determined that driving the thorium fuel cycle with low-enriched uranium is the most efficient way to minimize natural uranium consumption. Assuming that a 40 MWd/kg exit burnup was achieved in the CANDU reactors, the fuel cycle yielded an 82% savings of natural uranium, compared to a scenario in which all power came from PWRs, while a 20 MWd/kg exit burnup increased the savings to 94%. The savings ranged over those exit burnups from 55%-69% for the variant with PWR-derived plutonium, 60%-73% for PWR-derived MOX fuel, and 78%-87% for CANDU-derived plutonium. The thermal power, radioactivity, and health hazards of the spent fuel were the highest for the case with MOX fuel driver but americium recycling proved effective for decreasing the long term dangers. In a global reactor park, the introduction of the thorium fuel cycle was hampered by the availability of fissile resources and, for a PWR-derived driver, natural uranium consumption was only reduced by 22% over 100 years relative to the PWR only scenario.</p> / Master of Applied Science (MASc)