Part A of two parts. Part B titled: "Xenon Transient Studies For a CANDU Reactor". / <p> In CANDU reactors, mechanical devices called shutoff rods are used to shutdown the reactor if required. These rods are made of high thermal neutron absorbing material such as cadmium. The number and the locations of the shutoff rods are optimized for a given reactor configuration. Optimization here means minimizing the number of rods and maximizing their reactivity depth or effectiveness. </p> <p> Optimization may be studied in various ways but the method selected is both simple and basic. It is apparent that if the interaction effects between the individual shutoff rods are reduced, their worths will increase. The optimum distance between two rods was determined to be 130 cm. Also, the best location of a third rod with respect to two already placed at an optimum separation was studied. Finally, these results were used in order to determine the optimum distance between banks of shutoff rods. These banks of rods were arranged in such a way as to achieve maximum flux flattening with all the rods inserted in the core. A 22 shutoff rod configuration for an adjuster flattened CANDU reactor gave a total change of 5.6% in keff. </p> / Thesis / Master of Engineering (MEngr)

In the process of exploring the thorium fuel application in CANDU reactors, it is important to consider the proliferation resistance level as a parameter for comparison with current natural uranium fuel. The concept of a whole fuel cycle was introduced to show the variations in the proliferation resistance level as the material is flowing through the cycle. The depletion and decay histories were simulated with SCALE 6.1 code and the results such as isotopes composition, decay heat, and radioactivity were used to analyze the material attractiveness of pure heavy metal for weapon production. They also served as the intrinsic features during the proliferation resistance level calculation. The Multi-Attribute Utility Analysis (MAUA) method developed by Chalton was used to compare different CANDU fuel cycles with quantified values (PR) from the viewpoint of proliferation resistance. To improve the biased MAUA results that gave a PR of 0.76 to CANDU while 0.93 to PWR, the attributes of size/weight and refueling scheme were reconsidered. In addition, the sensitive technology involved was added for the proliferation resistance recalculation. The results showed an increased PR value of 0.82 for natural uranium CANDU reactor as well as a decreasing trend of PR at the back end. PWR has a PR of 0.82 with revised MAUA method. The PR comparison of thorium and natural uranium fuel indicated that Th/Pu fuel has a slightly higher PR value in the reactor. The Figure of Merit (FOM) method developed by Bathke was used to validate the PR results from another perspective: the attractiveness of pure heavy metals that are suitable for nuclear weapon production. The results showed that FOM of plutonium keeps increasing with decay time and the trend becomes more significant after disposal in the deep geological repository. The FOM of uranium from Th/Pu cases is higher than that of Pu within several hundred years but maintains a decreasing trend. The decreasing FOM of uranium is preferred for direct disposal in deep geological repository. The decreased PR level and the increased FOM value of plutonium at the back end of a fuel cycle indicate the importance of implementing the security and safeguard for each facility dealing with nuclear materials. The comparison results of PR and FOM values for different fuel provided feedback and suggestions for the new fuel application. / Thesis / Master of Applied Science (MASc)

One of two project reports: The other part is designated as Part A / <p> A fundamental and important part of nuclear reactor development and analysis today is the study of neutronics following a breach in the primary heat transport circuit. In the past, much of this analysis has concentrated on the calculation of the thermalhydraulic changes which occur following a loss of coolant accident and the effects these subsequently have on neutron kinetics. The purpose of this present study is to examine the influence of neutronic parameters on the size and shape of power pulses which result from loss of coolant accidents. The parameters studied are shutdown system delay times, shutoff rod drop curves, and fuel burnup distribution. </p> / Thesis / Master of Engineering (MEngr)

Flow in the Moderator of a CANDU reactor can be very complex due to the interplay of convective and buoyant effects. Experiments have been performed to measure temperature and velocity fields for these kind of flows, although concerns still exist. As a result a Moderator test facility has been built in order to validate CFD models for future predictions and safety analysis. To properly validate this experiment an accurate set of inlet flow conditions must be established in order to ensure a fair comparison. A series of flow conditions indicative of the header assemblies which feed flow into the moderator test facility have been investigated through experimentation, empirical evaluation and numerical simulation. They include flow through curved tubes, turbulent free jets and flow through dividing manifolds. The goal of the present study is to establish the modelling approach to predict the flow distribution inside the manifold and velocity field out of the J-nozzles. A variety of RANS based turbulence models and computational meshes were employed in the numerical study. The turbulence model that was found to perform best was the realizable k- model. It was also found that the velocity field of the J-nozzles is constant between Reynolds numbers of 6800-9300. These Reynolds numbers are indicative of those expected out of the header assemblies. / Thesis / Master of Applied Science (MASc)

Since the Fukushima Daiichi nuclear accident, there have been ongoing efforts to enhance the modelling capabilities for severe accidents in nuclear power plants. The primary severe accident analysis code used in Canada for its CANDU reactors is MAAP-CANDU (adapted from MAAP-LWR). In order to meet the new requirements that have evolved since Fukushima, upgrades to MAAP-CANDU have been made most recently by the Canadian nuclear industry. While the newest version (i.e. MAAP5-CANDU) offers several important improvements primarily in core nodalization and core collapse modelling, it still lacks mechanistic models for many key thermo-mechanical deformation phenomena that may significantly impact accident progression and event timings. It is also a general consensus that having alternative analysis tools is beneficial in improving our confidence in the simulation results, especially given the complex nature of severe accident phenomena in CANDU and the limited experimental support. This thesis seeks a novel approach to CANDU severe accident modelling by combining the best-estimate thermal-hydraulic code RELAP5, the severe accident models in SCDAP, and several CANDU-specific mechanistic deformation models developed by the author. This work mainly consists of two parts. The first part is focused on the assessment of natural circulation heat sinks following crash-cooldown in the early-phase of a Station Blackout (SBO) accident where fuel channel deformation can be precluded. The effectiveness of steam generator heat removal after crash-cooldown and that of the several water make-up options were demonstrated through the simulation of several SBO scenarios with/without crash-cooldown, sensitivity studies, as well as benchmarking against station and experimental measurements. In the second part, several mechanistic severe accident models were developed to enhance the simulation fidelity beyond the initial steam generator heat sink phase to the moderator boil-off and core disassembly phases. This includes models for predicting the pressure tube ballooning and sagging phenomena during the fuel channel heat-up phase and models for the sagging and disassembly of fuel channel assemblies during the core disassembly phase. After benchmarking against relevant channel deformation experiments, the models were successfully integrated into the RELAP/SCDAPSIM/MOD3.6 code as part of the SCDAP subroutines. The advantage of utilizing a code such as SCDAP is that generic models for fission product release and hydrogen generations, which are well benchmarked, can be directly applied to CANDU simulations. With the modified MOD3.6 code the early-phase SBO simulations were extended to include the later stages of SBO until the calandria vessel dryout. The current modelling approach replaced the simple threshold-type models commonly seen in the integrated severe accident codes such as MAAP-CANDU with more mechanistic models thereby providing a more robust treatment of the core degradation process during severe accident in CANDU. / Thesis / Doctor of Philosophy (PhD)

The IAEA ICSP on Numerical Benchmarks for Multiphysics Simulation of Pressurized Heavy Water Reactor Transients was initiated in 2016 to facilitate the development of a set of open access, standardized, numerical test problems for postulated accident scenarios in a CANDU styled Reactor. The test problems include a loss of coolant accident resulting from an inlet header break, a loss of flow accident caused by a single pump trip, and a loss of regulation accident due to inadvertently withdrawn adjusters. The Benchmark was split into phases, which included stand-alone physics and thermal-hydraulics transients, coupled steady state simulations, and coupled transients. This thesis documents the results that were generated through an original TRACE/PARCS coupling methodology that was developed specifically for this work. There is a strong emphasis on development methods and step by step verification throughout the thesis, to provide a framework for future research in this area. In addition to the Benchmark results, additional studies on propagation of fundamental nuclear data uncertainty, and sensitivity analysis of coupled transients are reported in this thesis. Two Phenomena and Key Parameter Identification and Ranking Tables were generated for the loss of coolant accident scenario, to provide feedback to the Benchmark Team, and to add to the body of work on uncertainty/sensitivity analysis of CANDU style reactors. Some important results from the uncertainty analysis work relate to changes in the uncertainty of figures of merit such as integrated core power, and peak core power magnitude and time, between small and large break loss of coolant accidents. The analysis shows that the mean and standard deviation of the integrated core power and maximum integrated channel power, are very close between a 30% header break and a 60% header break, despite the peak core power being much larger in the 60% break case. Furthermore, it shows that there is a trade off between the uncertainty in the time of the peak core power, and the magnitude of the peak core power, with smaller breaks showing a smaller standard deviation in the magnitude of the peak core power, but a larger standard deviation in when this power is reached during the transient, and vice versa for larger breaks. From the results of the sensitivity analysis study, this thesis concludes that parameters related to coolant void reactivity and shutoff rod timing and effectiveness have the largest impact on loss of coolant accident progressions, while parameters that can have a large impact in other transients or reactor designs, such as fuel temperature reactivity feedback and control device incremental cross sections, are less important. / Thesis / Master of Science (MSc) / This thesis documents McMaster’s contribution to an International Atomic Energy Agency Benchmark on Pressurized Heavy Water Reactors that closely resemble the CANDU design. The Benchmark focus is on coupling of thermal-hydraulics and neutron physics codes, and simulation of postulated accident scenarios. This thesis contains some select results from the Benchmark, comparing the results generated by McMaster to other participants. This thesis also documents additional work that was performed to propagate fundamental nuclear data uncertainty through the coupled transient calculations and obtain an estimate of the uncertainty in key figures of merit. This work was beyond the scope of the Benchmark and is a unique contribution to the open literature. Finally, sensitivity studies were performed on one of the accident scenarios defined in the Benchmark, the loss of coolant accident, to determine which input parameters have the largest contribution to the variability of key figures of merit.

Fuels comprised of the element thorium have become increasingly popular with researchers and the public as the next generation fuel due to its ability to produce its own fissile element (U-233) and generate lower concentrations of heavy actinides. The use of thorium can possibly lead to a self-sustaining cycle whereby the addition of fissile material is not required and that the fuel can breed sufficient amounts of U-233 for a continuous supply. Research into thorium use in CANDU reactors has mainly been focused on using driver elements such as U-235 or Pu-239 to initiate the nuclear reaction by taking advantage of bundle design or by mixing the thorium and driver fuel together; however, these methods have added complexities and may not lead to a pure thorium fuel cycle, but extend the life of current nuclear fuels used. This thesis will investigate a simpler means of utilizing thorium for the intent of breeding U-233 through the use of pure thorium bundles in a once-through cycle by the ways of a heterogeneous core loading in a CANDU-6 reactor model. A 3x3 multi-cell model using DRAGON 3.06K will simulate the dual fuel model by having the centre lattice enclosing the thorium bundle and the outer eight lattices enclosing the enriched uranium bundles as the driver fuel. Next, the diffusion code DONJON 3.02E is used to produce time-average, instantaneous, and initial startup full-core simulations. As well, a brief look at the refuelling operations on the thorium channels will be done. The presence of a thorium bundle places a negative reactivity load on the multi-cell, but causes a positive insertion of reactivity for a coolant void and shutdown scenario. In the full-core modelling, the final core configuration chosen shows that thorium channels should be located in the inner core rather than in the most outer channels to produce a flattening effect on the radial profile. Thorium channels will require a combination of SEU and thorium bundles in an attempt to maintain channel power levels. Specifically, the use of 4, 6, or 8 Th bundles were investigated. The most optimal core performance shown has a radial form factor of 0.816, a total average core burnup of 18.32 GWd/t, and operates within designed power limits. It is possible to implement pure thorium bundles into a reactor set in a dual fuel mode. A careful consideration of where thorium bundles should be located in the core can help flatten the radial power distribution and help the reactor operate within the operating licensing parameters without the use of adjuster rods while breeding U-233 for a future thorium fuel cycle. / Thesis / Master of Applied Science (MASc)

Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort. / UOIT

The National Fire Code of Canada (NFCC) is one model code which regulates hot work in Canada. The code specifies that hot work processes need only create heat to be considered hot work processes, and requires that precautions taken adhere to those in Canadian Standards Association (CSA) W117.2, which is intended for welding, cutting and allied processes. CSA W117.2 requires a 15 m spherical radius of separation in which combustibles are ideally relocated or, at minimum, be protected with fire blankets. Openings, cracks and other locations in which sparks or hot particles must also be protected within this distance. Additionally CSA W117.2 requires a fire watch during, and one hour following the completion of the work. The NFCC stipulates more stringent requirements on the fire watch than CSA W117.2, requiring a check back 4 hours after the work. The code in its current form requires the same precautions be taken when using a soldering iron or epoxy resin as when using an oxyacetylene torch to flame cut steel. The lack of hazard characterization of hot work processes, and the umbrella prescription of required fire safety precautions can result in insufficient measures to prevent fires in some scenarios, and inordinate precautionary measures in others. While not applicable law in all jurisdictions, the NFCC is relied on in various Canadian industries for regulative purposes. Nuclear power generation in Canada is one such industry facing onerous fire protection costs resulting from following these precautions for the smallest of jobs requiring heat producing tools. The literature review highlights the dearth of scientific knowledge regarding the propensity of hot work as an ignition source and how this shortcoming manifests itself in issues across the various standards governing hot work practices. The objective of this research is to assess fire hazards resulting from various processes considered hot work under the National Fire Code of Canada (NFCC). Due to the breadth of processes covered by the NFCC, a spectrum of hot work activities was investigated from processes as innocuous as the application of heated adhesive, to well known sources of ignition such as a variety of welding processes, oxyacetylene cutting and plasma cutting. To streamline the hazard assessment, processes were categorised into three groups based upon expected hazards such that testing could focus on the most prominent ignition danger presented by each. The groups were those processes exhibiting hot surface ignition hazards, processes with hot surface ignition hazards in addition to limited potential to generate hot particles, and those processes in which the generation of significant quantities of spark and hot particles is guaranteed. For the first two process categories, experimentation focused on determining a critical process temperature with which to rank processes and also compare with ignition temperatures of combustibles commonly involved in hot work fires. The critical process temperature was determined as the highest measured temperature of the workpiece or tool during the chosen process and was typically measured with the use of thermocouples and infrared thermography. Characterization of any hot particles in the second category was performed using infrared thermography, and in some cases, thermal paper. Literature sources indicated that sparks and hot particles are the largest factor in hot work fires, so specialised methodology was developed for the third category of processes to characterise the distribution of many thousands of hot work particles generated during welding, thermal cutting and other hot particle producing work. The distributions collected were used to determine the area enveloped by the ignition hazard of hot particles as well as areas encompassing the highest threat to combustibles in relative terms. Several of the processes as studied were found not to exhibit any measurable form of ignition hazard, including forms of manual sanding and filing and rotary filing of steel. Heated adhesive, cutting steel with a reciprocating saw and drilling of steel were shown to exhibit moderate degrees of hazard with temperature rise of 195\degree C or less, suggesting potential hazard to a limited group of combustibles. Welding and cutting processes were shown to have a relative ignition potential across a wide area. Typical welding procedures produced hot particles which travelled a maximum of approximately 3 - 4 m while thermal cutting processes ejected sparks, slag and hot particles up to 9.8 m from the work. Incorporated properly into updated standards and codes, the results and findings of this research could drastically improve the Canadian model codes regarding the regulation of hot work by decreasing cost and difficulty for Canadian Industry without increasing the risk of loss.

Thesis (Ph.D.) -- McMaster University, 1999. / Includes bibliographical references (leaves [188]-192). Also available via World Wide Web.