A comparison of available fluid-to-fluid scaling laws for scaling convective heat transfer at supercritical pressures showed that the ones suggested by Zahlan, Groeneveld and Tavoularis (ZGT) have some advantages. The applicability of the ZGT laws was tested for pairs of fluids including carbon dioxide, water or Refrigerant R134a. The conditions of previous measurements taken in the Supercritical University of Ottawa Loop with CO2 flowing vertically upwards in an electrically heated tube with 8 mm ID were scaled to equivalent conditions in R134a and new measurements of the heat transfer coefficient (HTC) were taken in the same tube using the latter fluid. The inlet pressure was 1.13 times the critical pressure (4.06 MPa), the mass flux was in the range from 212 kg/m^2 s to 1609 kg/m^2 s, the heat flux was in the range from 2 kW/m^2 to 137 kW/m^2, and the inlet temperature was in the range from 62 ℃ to 105 ℃. The HTC at equivalent conditions in water was also determined with the use of transcritical look-up tables. Average and linearly varying corrections to the ZGT scaling laws were derived by statistical analysis for each pair of fluids under NHT or DHT conditions. Such corrections reduced the standard deviation of the scaling error but did not eliminate the presence of large errors under many sets of conditions. As expected, scaling errors were in general larger for DHT than NHT conditions. The present results did not reveal any systematic and correctable dependence of the scaling error upon the mass flux or heat flux but showed that scaling errors became particularly large as the bulk temperature T_b approached the pseudocritical temperature T_pc. In conclusion, the ZGT scaling laws appear to be fairly accurate for the three pairs of fluids considered in the liquid-like region with T_b/T_pc ≤ 0.94 and possibly in the gas-like region with T_b/T_pc ≥ 1.02, whereas outside this range scaling errors could be significant. It was also found that the ZGT scaling laws do not scale accurately the onset of DHT in different fluids.

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 reactor‘s high outlet temperatures. A generic pressure-channel (or pressure-tube)SCWR operates at a pressure of 25 MPa with inlet- and outlet-coolant temperatures of 350°C and 625°C. Consequently, the sheath and fuel centerline temperatures are higher in SCWRs than those of the current nuclear reactors. Previous studies have shown that the sheath and fuel centerline temperatures could exceed the design and industry accepted limits of 850°C and 1850°C, respectively. These studies correspond to UO2 enclosed in a 43-element fuel bundle at an average thermal power per channel of 8.5 MWth. Additionally, these high operating conditions in the range of 350 - 625°C lead to high heat losses from the coolant to the moderator, which in turn reduces the overall thermal efficiency of the Nuclear Power Plant (NPP). Therefore, there is a need for alternative fuels or fuel bundles for future use in SCWRs. Hence, it is also necessary to determine the amount of heat losses from a number of fuel-channel designs for SCWRs. The objectives of this study are to investigate the possibility of using alternative fuels and to determine the heat losses from a fuel-channel design at SCWR conditions. The investigated fuels are categorized as low thermal-conductivity (e.g., UO2, MOX, and ThO2), high thermal-conductivity (e.g., UC, UC2, UN), and enhanced thermal-conductivity (e.g., UO2‒SiC, UO2‒C, and UO2‒BeO) fuels. Additionally, the examined fuel channel is the High Efficiency Channel (HEC), which has been designed by the Atomic Energy of Canada Limited (AECL) for the proposed CANDU SCWR. In order to achieve the objectives of this study, a steady-state one-dimensional heat-transfer analysis was conducted. The MATLAB© and NIST REFPROP© software were used for programming and retrieving thermophysical properties of a light-water coolant, respectively. The fuel centerline temperature was calculated for the fuel channels with the maximum thermal power, i.e., +15% above average channel power. Results of this analysis showed that the fuel centerline temperatures of low thermal-conductivity fuels exceed the industry limit; therefore, either a fuel with a higher thermal conductivity should be used or the fuel bundle geometry must be modified. Among the high thermal-conductivity fuels, UC has been shown to be a candidate for future use in SCWRs. However, the chemical compatibility of UC with water at high operating temperatures of SCWRs remains ambiguous. Therefore, further studies are required before selecting UC. In regards to enhanced thermal-conductivity fuels, UO2‒BeO is the most suitable candidate; however, its mechanical and neutronic properties must be thoroughly studied before any decision is made with regards to the selection of a fuel. In regards to the heat losses from the examined fuel channel, the heat loss was between 70 kW and 110 kW per fuel channel based on an average thermal power per channel of 8.5 MWth and a moderator pressure of 0.1 MPa at 80°C. A sensitivity analysis of the fuel channel shows that the heat loss can be reduced by increasing the operating pressure of the moderator, which in turn allows for increasing the operating temperature of the moderator. Higher operating temperatures of the moderator result in smaller temperature differences between the coolant and the moderator, which leads to lower heat losses. Therefore, either the thickness of the insulator or the pressure of the moderator should be increased in order to reduce the heat losses from the fuel channel. / UOIT

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

The aim of this work was to rationalize the corrosion resistance of candidate austenitic iron-chromium-nickel (Fe-Cr-Ni) alloys in supercritical water (SCW) for use as fuel claddings within the in-core structure of the Canadian supercritical water-cooled reactor (SCWR) concept. High chromium (Cr)-containing alloys (Alloy 800HT with 20.6 wt.% Cr and 30.7 wt.% Ni and Alloy 33 with 33.4 wt.% Cr and 31.9 wt.% Ni) in the mill annealed condition were chosen for this purpose. Coupons were exposed on a short-term basis (500 h) in a static autoclave containing 25 MPa SCW at 550 °C and 625 °C. Gravimetric measurements and electron microscopy techniques were then used to study the oxidation/corrosion resistance of two alloys. Alloy 33 was found to exhibit the higher corrosion resistance at both temperatures. The improved corrosion resistance of Alloy 33 was attributed to two factors: (i) the formation of a continuous Cr-rich corundum-type M2O3 (M= Cr and Fe) oxide layer that prevented the diffusion of Fe and the formation of a less-protective Fe/Mn-Cr spinel ([Fe,Mn]Cr2O4) outer layer, (ii) a sufficient residual bulk Cr in the Cr-depleted layer adjacent to the alloy/scale interface that prevented any localized internal oxidation from occurring. A mass balance conducted on the corroded Alloy 33 material suggested that volatilization of the corundum-type oxide layer did not occur, at least not within the short-term exposure in the essentially deaerated SCW. A key issue requiring further study was the observation of intermetallic precipitates that formed below the Cr-depleted zone adjacent to the alloy/scale interface in both alloys when exposed for 500 h at 625 °C and their possible influence on the in-service mechanical integrity. / Thesis / Master of Applied Science (MASc) / The supercritical water-cooled reactor (SCWR) is one of the six reactor design concepts developed by the Generation-IV International Forum (GIF). Canada is planning to build the SCWR within the next decades. However, selection of proper materials that perform well within such high pressure high temperature circumstances inside the reactor core with minimum degradation is a very imperative challenge. The current work has addressed this issue by studying the corrosion behaviour of Fe-Cr-Ni alloys in similar environment using electron microscopy techniques.

The Pressure Tube Super Critical Water Reactor (PT-SCWR) is a design with a light water coolant operating at 25 MPa above the thermodynamic critical pressure, with a separated low pressure and temperature moderator, facilitated by a High E ciency Channel consisting of a pressure tube and a porous ceramic insulator tube. The 2011 AECL reference design is considered along with a 2012 benchmark. In the 2011 reference design the coolant is permitted to ow through the insulator. The insulator region has a temperature gradient from 881 K at the inner liner tube to 478 K at the pressure tube wall. The density of light water varies by an order of magnitude depending on the local enthalpy of the uid. The lateral coolant density is estimated as a radial function at ve axial positions with the lattice physics codes WIMS-AECL and Serpent. The lateral coolant density variations in the insulator region of the PT-SCWR cause strong reactivity and CVR e ects which vary heavily on axial location due to the changes in the estimated mass of coolant and the physical relocation of the coolant closer to the moderator, as the coolant is estimated to be least dense closer to the fuel region of the coolant ow. The beta version of Serpent 2 is used to explore the lateral coolant densities in the subchannel region of the insulator in the 2012 version of the PT-SCWR. A more advanced coolant density analysis with FLUENT is used to estimate the subchannel coolant density variation, which is linked to SERPENT 2s multi-physics interface, allowing the lattice code to measure the sensitivity of the model to the analysis of the subchannels. This analysis increases the reactivity of the PT-SCWR through the displacement of the coolant. Serpent 2 is accepted as a valid lattice code for PT-SCWR analysis. / Thesis / Master of Applied Science (MASc)

Knowledge of the effects of nuclear data uncertainties and physics approximations is crucial for the development, design, operation, and accident mitigation, of nuclear power plants. A framework to create a simulated fuel bundle, based on sensitivities and similarities, has been developed. The methodology allows safe-to-handle fuel to be manufactured such that it mimics irradiated fuel and can be used to reduce simulation uncertainties and better predict an application’s response. In this work, similarity values of ck = 0.967, E = 0.992, and G = 0.891 were found between between the irradiated fuel, and non-irradiated simulated fuel. In addition, a set of ZED-2 experiments has been analyzed that are applicable to an SCWR nuclear data adjustment and simulation bias determination. This was shown through high sensitivity coverage of many important nuclides, however, a low completeness value of R=0.24 indicates the set of 39 experiments alone is not sufficient for an accurate bias determination. Lastly, a technique has been presented that reduces diffusion calculation errors through the use of novel and practical mean discontinuity factors. The discontinuity factors have shown to reduce maximum channel power errors by up to 6.7%, and reactivity errors by 2.6 mk, compared to conventional analysis techniques. / Thesis / Doctor of Philosophy (PhD) / Use of practical discontinuity factors has shown to reduce channel power predictions significantly. Furthermore, an experimental and numerical technique has been developed to improve neutron transport predictions. Finally, a set of experiments have been modeled and simulated to determine their applicability to the SCWR.

<p>In the next few decades, the nuclear industry worldwide is expected to launch a set of reactors with advanced reactor designs. Generation-IV (GEN-IV) reactors are to display superior safety by incorporating additional passive safety concepts as well as improving accident management and minimization of consequences. Canada is in the early stages of conceiving its GEN-IV reactor design – the Supercritical Water Reactor (SCWR). The proposed design is based on the existing CANDU configurations and is expected to offer significant advances in thermal efficiency, fuel cycle sustainability, and relative cost of energy. Of particular interest is the reactor's ability to use inherent or passive safety concepts which will translate to the reactor being walk-away safe in an accident.</p> <p>Steam generators in CANDU remove decay heat by thermosyphoning in a loss of Class-IV power accident. This natural circulation process was a passive feature in GEN-II and GEN-III CANDUs. The SCWR's direct thermodynamic cycle implies steam generators are no longer incorporated into the design. This thesis examines how the SCWR compensates for the removal of a passive safety system element and the difference to the overall safety of the reactor following accidents. These results will be compared to the traditional CANDU's response in accidents to demonstrate the added value of this new reactor in maintaining the goal of no widespread core damage. Comparisons were also made between the SCWR and similar GEN-IV reactors in terms of design and response to various initiating events.</p> <p>Probabilistic Risk Analysis is used in this thesis to assess the SCWR design options. Although the SCWR is in the pre-conceptual design phase, the results of such risk assessment studies could affect the design, operation, and licensing of this new reactor. Future studies can build on this work to conduct more detailed analyses to characterise the SCWR's safety and reliability.</p> / Master of Applied Science (MASc)

<p>Among the six GEN-IV reactor concepts recommended by the Gen-IV International Forum, supercritical water-cooled reactors (SCWR) have gained significant interests due to its economic advantage, technology and experience continuity. In the last few years, extensive R&D activities have been launched covering the various aspects of SCWR development, especially in thermal-hydraulic analysis. In Canada, most R&D projects are led by AECL or NRCan.</p> <p>SCWR design and development require the modification of simulation codes used for design and safety demonstration of subcritical water-cooled reactors. This study modifies the subchannel code COBRA-TF, applicable to only subcritical water-cooled reactors, to a new version COBRA-TF-SC, applicable to both supercritical and subcritical water-cooled reactors. Supercritical water property data tables and supercritical water property formulations are implemented. Supercritical water heat transfer and pressure drop correlations are also added. The saturation curve in the subcritical model is extended by introducing a pseudo two-phase region at supercritical pressures to avoid any numerical instabilities consistent with other studies.</p> <p>Some simple fuel bundle experimental data on the flow and temperature distribution are used to evaluate the code. The fuel bundle experiment is simulated with both COBRA-TF-SC and AECL's ASSERT-PV-SC. The COBRA-TF-SC predicted results show good agreement with the experimental data and results obtained from ASSERT-PV-SC, demonstrating good feasibility and accuracy of this code. COBRA-TF-SC is then used to predict the detailed thermalhydraulics behaviour of the 62-element Canadian SCWR fuel bundle design. The advantage of COBRA-TF-SC is that it can accommodate transcritical flow conditions whereas the existing subchannel codes for SCWRs cannot.</p> / Master of Applied Science (MASc)

The Generation IV International Forum (GIF) has initiated an international collaboration for the research and development of the Generation IV future nuclear energy systems. The Canadian PT-SCWR is Canada’s contribution to the GIF as a GEN-IV advanced energy system. The PT-SCWR is a pressure tube reactor type and considered as an evolution of the conventional CANDU reactor. The PT-SCWR is characterized by bi-directional coolant flow through the High Efficiency Re-entrant Channel (HERC). The Canadian SCWR is a unique design involving high pressure and temperature coolant, a light water moderator, and a thorium-plutonium fuel, and is unlike any operating or conceptual reactor at this time. The SCWR does share some features in common with the BWR configuration (direct cycle, control blades etc…), CANDU (separate low temperature moderator), and the HTGR/HTR (coolant with high propensity to up-scatter), and so it represents a hybrid of many concepts. Because of its hybrid nature there have been subtle feedback effects reported in the literature which have not been fully analyzed and are highly dependent on these unique characteristics in the core. Also given the significant isotopic changes in the fuel it is necessary to understand how the feedback mechanisms evolve with fuel depletion. Finally, given the spectral differences from both CANDU and HTR reactors further study on the few-energy group homogenization is needed. The three papers in this thesis address each one of these issues identified in literature. Models were created using the SCALE (Standardized Computer Analysis for Licensing Evaluation) code package. Through this work, it was found that the lattice is affected by more than one large individual phenomenon but that these phenomena cancel one another to have a small net final change. These phenomena are highly affected by the coolant properties which have major roles in neutron thermalization process since the PT-SCWR is characterized by a tight lattice pitch. It was observed that fresh and depleted fuel have almost similar behaviour with small differences due to the Pu depletion and the production of minor actinides, 233U and xenon. It was also found that a higher thermal energy barrier is recommended for the two-energy-group structure since the PT-SCWR is characterized by a large coolant temperature compared to the conventional water thermal reactors. Two, three and four optimum energy group structure homogenizations were determined based on the behaviour of the neutron multiplication factor and other reactivity feedback coefficients. Robust numerical computations and experience in the physics of the problem were used in the few-energy group optimization methodology. The results show that the accuracy of the expected solution becomes highly independent of the number of energy groups with more than four energy groups used. / Thesis / Doctor of Philosophy (PhD)