Inconel alloys are commonly used as structural materials in nuclear reactors. One of these alloys, the Inconel X-750, is a γ’ Ni3(Al, Ti) strengthened superalloy extensively used in the cores of reactors, such as spacers in CANada Deuterium Uranium (CANDU) fuel channels. Prior to their application in commercial reactors, accelerated irradiation tests had been conducted in liquid metal fast reactors. Results did not indicate any problem stemming from significant fast neutron irradiation. However, recently it has been found that the ex-service CANDU Inconel X-750 spacers became severely brittle after a lengthy exposure to reactor environment. The underlying mechanism remains unclear and thus forms the focus of this current investigation, predominantly through transmission electron microscopy (TEM). This dissertation unfolds with the literature review in Chapter 2, followed by presentation of novel techniques in Chapter 3 on the preparation of TEM samples from small reactor components, namely the spacers. Chapter 4 presents TEM characterizations of ex-service spacers removed from the reactors. To simulate neutron irradiation over wide temperature range in an effort to understand the damage mechanisms, heavy ion irradiations were conducted and reported in Chapter 5 and 6. Irradiations are found to significantly alter the stability of the primary strengthening phase γ’, a systematic experimental study of which is presented in Chapter 7. To fully understand the effects of transmutation produced helium on irradiation induced cavity and dislocation microstructures, TEM in-situ heavy ion irradiations with hot/cold pre-injected helium were conducted and reported in Chapter 8 and 9. Helium was found to play an important role in the irradiation-induced instability of γ’ in nickel-based superalloys, the discussion of which is presented in Chapter 10. As one of the most important defect structures induced from irradiation, the stacking-fault-tetrahedra, were dynamically observed and are described in a journal manuscript in Appendix A. In addition to broadening current understanding of material degradation mechanism for in-service CANDU spacer, this study also provides comprehensive information on irradiation damage in nickel based superalloys, irradiation induced lattice defects and phase instability in face centered cubic alloys, as well as helium’s effects on cavity formation, dislocation evolution, and phase transformation. / Thesis (Ph.D, Mechanical and Materials Engineering) -- Queen's University, 2013-09-06 15:21:02.334

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)