The moderator within the CANDU calandria vessel is an important element of the reactor serving various operational and safety functions. A validated computational fluid dynamics model to simulate the fluid within the calandria vessel is required to predict how it would behave in different situations. Given is high level of complexity (flow from complicated inlets inlet, outlet, volumetric heating, flows around fuel channels, etc…) validation typically proceeds from separate effect studies for each component as well as integral validation where all complex phenomena are included. This work is focused on the separate effect studies aimed at phenomena relevant to the inlets of Pickering A. A scaled down model of the Pickering CANDU moderator inlet nozzle was constructed and used to verify simulations done on the same nozzle. Physical measurements were taken using a PIV laser system on the scaled down nozzle. Seven measurements planes were taken on two of the outlets of the nozzle at two different flowrates (0.2 kg/s and 0.1 kg/s). Given a single PIV measurement plane contains approximately 1000 velocity points this represents approximately 28000 new velocity measurements for CFD validation. A CAD model of the nozzle was imported into STAR-CCM+ CFD software to perform the simulations. The simulations used the k-ω SST turbulence model and were performed using the same flowrates as in the experiments. The k- model was also used as a comparison for the k- SST model. The same locations were used in calculations for the seven measurement planes so that the same areas could be compared for the models employed. The CFD results were found to agree qualitatively with the measured ones but no quantitative comparisons should be made with care. Additional investigation is required for a more comprehensive comparison. While both, measured and calculated results, showed similar velocity distributions, the velocities values differed by varying degrees. In particular, the simulations did not show the same jet expansion as the physical measurements indicated. There is not enough data from the measurement planes to determine if the flow in the physical experiments are more diffuse and the velocity field gradients are lower than obtained in the current study. With large portions of the nozzle unobserved by the measurement planes, it cannot be concluded that there are no other peaks or large velocity gradient regions in between the planes. For future work it would be beneficial to have a larger model of the nozzle for better gradient resolution, as well as using a measuring system that allows for more precise data collection. Different turbulence models and measurements further away from the nozzle should also be used to determine which would be best in a larger simulation. / Thesis / Master of Applied Science (MASc) / In order to create accurate simulations moderator flow in a calandria vessel, simulations of smaller components must first be verified. The moderator inlet fan nozzle found in the Pickering CANDU calandria vessel has complicated geometry which can be difficult to model using CFD. Simulations were performed on a scaled down version of the nozzle and physical experiments were performed on a constructed model. The results found that the simulations predicted similar velocity distributions, but generally with higher peak velocity values than the physical experiments. However additional measurement locations would be needed to give a more comprehensive comparison. Going forward with the larger simulations, to better determine the best model to use for the nozzle, a larger version should be used with different turbulence models and measurement locations.

The prediction of the moderator temperature distribution in a CANDU reactor is important in establishing its ability to act as an emergency heat sink for certain beyond design basis accidents. This analysis typically relies on computational models which are benchmarked against experimental data from small-scale test facilities. These small-scale models prioritize the matching of the Archimedes number (Ar) of the full-scale reactor, which represents the ratio of buoyancy forces to inertial forces. Concerns regarding similarity between the reactor and small-scale facilities may exist due to a large difference in scale, as well as geometric simplifications made due to practical limitations. This study examined the behaviour within an approximately 1/16 scale facility representative of the Bruce A calandria vessel, which features a unique inlet and outlet configuration. Experimental measurements were obtained for a range of power and flow conditions. Unsteady RANS simulations of the small-scale facility were also performed using the realizable k-epsilon model. Goals of the study included the assessment of the unique moderator inlet on the flow patterns inside the calandria vessel and how well existing CFD modelling approaches replicated these features. The observed flow and temperature distributions in the scale facility did not appear highly sensitive to changes in Ar. For all tested conditions, a large front-to-back recirculation pattern resulted from the asymmetric inlet arrangement. Peak temperatures consistently occurred toward the front of the vessel where inertial flows were assisted by buoyancy induced flows. Under steady-state conditions, unsteady and three-dimensional behaviour was observed within the vessel. Temperature fluctuations near the upper rear end of the vessel arose from the unstable interaction between cool downward flow from the inlets and upward buoyant flow from the tube bank. In the peak temperature regions, flow direction was relatively consistent in the upward direction. The simulations tended to overpredict the peak temperatures within the vessel by approximately 0.5 – 3.8 ºC. This behaviour was attributed to the model tending to underpredict the upward velocities entering the base of the tube bank in the peak temperature regions. As Ar increased and buoyancy effects became more significant in determining the local velocities, agreement between the predicted and measured velocities was improved. The similarity between the small-scale model and the full-size reactor was also assessed through comparisons to existing simulations of the full-size calandria. There was qualitative similarity between the two geometries, albeit at lower Ar for the small-scale facility. This suggested that buoyancy effects were more significant in the small-scale facility compared to the full-size calandria. This was attributed to the use of surface heating (as opposed to volumetric heating in the reactor), and relatively high surface heat fluxes caused by a reduced number of tube bank elements. / Thesis / Doctor of Philosophy (PhD)

Detection systems used to monitor reactor operations are of significant interest as tools for verification of operator declarations. Current reactor site safeguards are limited to visual inspections and intrusive monitoring systems. The recent development of antineutrino detectors may soon allow real-time monitoring from an unobtrusive location. Antineutrinos are produced through beta decay of fission products in the core. The lack of charge and small mass of the antineutrino ensures an extremely low interaction probability with all matter, effectively making the particle impossible to shield. As the fuel isotopic composition changes with burn-up, the primary fission source changes from ²³⁵U to ²³⁹Pu. Since differing antineutrino energy spectra are produced by each fissionable isotope, the antineutrino flux will also change as a function of burn-up. Supported by reactor simulations from nuclear codes, antineutrino detectors may provide a window into the reactor core and provide inspectors with tools to verify legitimate operations. This thesis is focused on the antineutrino rate produced by CANadian Deuterium Uranium reactors (CANDU) during startup. A CANDU fuel bundle model was created with the TRITON module from the SCALE6.1 code to calculate isotopic antineutrino rates for a single bundle. A full core CANDU model that incorporates refueling was also created for the first 155 days of operation after startup by using a Python 2.6 script to handle pre- and post-calculations. All simulations were calculated using operational data from Point Lepreau Generating Station produced by proprietary codes for the forthcoming fresh core startup. Dependence of the antineutrino rate on power and bundle replacement was analyzed, with a ±10% change in power causing a ±10% change in antineutrino rate, and the CANDU detector effectively measuring a 10% decrease in power within 9 hours of collection time. Bundle refueling was shown to only slightly modify the antineutrino rate, requiring a target volume more than 20 times larger than the present detector to effectively identify the change due to the bundles refueled over a one week period. Diversion of 15% or more of the total amount of bundles can be effectively measured by the CANDU detector within a one month counting period. / Graduation date: 2012

This thesis documents significant contributions to the quantification of input and modelling uncertainties in the simulation of nuclear power plants. This work is intended to support the simulations that are performed to demonstrate the safety of nuclear power plants in general, and in CANDU reactors specifically. The work presented in this thesis extends the methodologies for uncertainty propagation established internationally to CANDU plants and pioneers the integration of these tools with important plant features in CANDUs, such as online fueling. This thesis documents a series of simulation studies performed to quantify the impact of uncertainties (primarily nuclear data uncertainties), on simulations of CANDU stations and light water reactors (LWRs). The novel part of this work includes quantifying the role of operational feedbacks such as online refuelling and reactor control systems, and important modelling uncertainties, on CANDU simulations. To achieve this objective, this thesis examines 4 important areas as documented in journal papers. To demonstrate understanding of the tools developed for the UAM-LWR benchmark and to support the ongoing international effort, select studies from the UAM-LWR benchmark study exercises were performed and published in the first journal paper. Time-dependent PWR neutronics exercises, considering both nuclear data and manufacturing uncertainties, were completed. This work found that the relative importance of nuclear data uncertainties and manufacturing uncertainties depended on whether the parameter of interest was “local”, such as pin power factors, or “global”, such as homogenized assembly properties. The second publication in this thesis documents the adaption of the tools from the first paper to consider CANDU specific features, such as spatial control systems and online refuelling. This paper demonstrated the significant effect that consistent feedback from fuelling operations has on reducing the total uncertainty in core level simulations of CANDU plants. The tools developed for this work were used to support downstream studies by generating extensive sets of realistic initial conditions for many different possible nuclear datasets. The next publications utilized the tools developed above and then extends the methods to include operational aspects of CANDUs in the assessments for the first time. In the third paper these methods were then used to demonstrate the tools’ capabilities to simulate an operational transient (a power maneuver from 100% full power to 59% full power) in a CANDU station and compared the resultant prediction and uncertainties to measure plant responses. A further study, on the role of nuclear data and initial burnup distribution uncertainty on a CANDU plant’s response to perturbations to liquid zone controller levels, was also performed to examine the effect of the commonly used “superposition principle” utilized in industry to make safety analysis of CANDU’s various fueling states more tractable. In both cases the role of nuclear data uncertainties was generally found to be similar in magnitude to the role of uncertainty in the core initial conditions. The results of this work support the continued safe operation of CANDU nuclear generating stations in Canada by quantifying the role of select uncertainties on safety simulation outputs, informing future BEPU analysis for CANDU plants and demonstrating the exceptional flexibility of the CANDU reactor design. This is reflected in one of the major conclusions of these works, which demonstrates that the natural feedbacks in CANDU operation help to minimize the effect of uncertainties in the outcome of many safety analysis. / Thesis / Candidate in Philosophy

The present work focuses on the study of two-phase flow behavior called “Intermittent Buoyancy Induced Flow” (IBIF) resulting from the loss of coolant circulation in a CANDU nuclear reactor core. The main objectives are to study steam bubble formation and migration through the pressure tube and into the feeder tubes and headers, and to study the effect of pressure tube sagging on the two-phase flow behavior during IBIF. Experiments are conducted using air and water flow at atmospheric pressure to qualitatively examine the IBIF phenomena. The test showed oscillating periodic behavior in the void fraction as the air vents. In addition to this, a mathematical model based on a simplified momentum balance for the liquid and gas phases was formulated. The model was further solved and compared to the experimental data. The model predictions showed a reasonable agreement within the investigated range of void fractions.

The present work focuses on the study of two-phase flow behavior called “Intermittent Buoyancy Induced Flow” (IBIF) resulting from the loss of coolant circulation in a CANDU nuclear reactor core. The main objectives are to study steam bubble formation and migration through the pressure tube and into the feeder tubes and headers, and to study the effect of pressure tube sagging on the two-phase flow behavior during IBIF. Experiments are conducted using air and water flow at atmospheric pressure to qualitatively examine the IBIF phenomena. The test showed oscillating periodic behavior in the void fraction as the air vents. In addition to this, a mathematical model based on a simplified momentum balance for the liquid and gas phases was formulated. The model was further solved and compared to the experimental data. The model predictions showed a reasonable agreement within the investigated range of void fractions.

The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The thesis presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower bound approach.

The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The thesis presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower bound approach.

The Korea Hydro and Nuclear Power has built a new modular type of dry storage facility, known as MACSTOR KN-400 at Wolsong reactor site. The building has the capacity to store up to 24000 CANDU spent fuel bundles in a 4 rows by 10 columns arrangement of silos. The MACSTOR KN-400 consists of 40 silos; each silo has 10 storage baskets, each of which can store 60 CANDU spent fuel bundles. The development of an effective method for spent fuel verification at the MACSTOR KN-400 storage facility is necessary in order for the International Atomic Energy Agency (IAEA) to meet with safeguards regulations. The IAEA is interested in having a new effective method of re-verification of the nuclear material in the MACSTOR KN-400 dry storage facility in the event of any loss of continuity of knowledge, which occasionally happens when the installed seals fail. In the thesis work, MCNP models of central and corner structures of the MACSTOR KN-400 facility are developed, since both have different types of re-verification system. Both gamma and neutron simulations were carried out using the MCNP models developed for MACSTOR KN-400. The CANDU spent fuel bundle with discharge burnup of 7.5 GWD/t (burned at specific power of 28.39 MW/t) and 10 years cooled was considered for radiation source term estimation. For both the structures, MCNP simulations of gamma transport were done by including Cadmium-Zinc-Telluride (CZT) detector inside the re-verification tube. Gamma analyses for different spent fuel bundle diversion scenarios were carried out. It was observed that for diversion scenarios wherein the bundles are removed from the inner portions of the basket (opposite side of the collimator of the re-verification tube), it was difficult to conclude whether diversion has taken place based on the change in gamma radiation signals. Similar MCNP simulations of neutron transport were carried out by integrating helium-3 detector inside the re-verification tube and the results obtained for various diversion scenarios were encouraging and can be used to detect some spent fuel diversion cases. In the central structure, it was observed that addition of moderating material between the spent fuel and the detector increased the sensitivity of the detecting system for various diversion cases for neutron simulations. In the worst scenario, the diverting state could divert 14 spent fuel bundles from each of 10 baskets in a silo from the basket region opposite to the collimator of the re-verification tube. The non-detection probability for this scenario is close to 1. This diversion cannot be easily detected using the currently designed detection system. In order to increase the detection probability, either the design of the facility must be changed or other safeguard methods, such as containment and surveillance methods must be used for safeguarding the nuclear material at the facility.

<p> The optimal placement of shutdown systems in power reactors is investigated, in particular, the placement of mechanical and liquid shutoff rods. Two CANDU reactor cores were used as a basis for evaluation. The optimal shutdown system was defined here to be one which, with the least number of rods, maximizes the reactivity depth of the system with the two most effective rods assumed to be absent. It was found that rows of rods placed parallel to the fuel channels were more effective and four of these rows were required in a simple core. For real cores where positions are limited six or seven rows were needed to obtain a large system worth. (Time analysis was not done to evaluate insertion rate and delay effects on the power transient in the case of an accident.) </p> / Thesis / Master of Engineering (MEngr)