This Study is performed to assess the Canadian pre-conceptual SCWR design for Loss of Flow events leading to degraded cooling conditions and to assess the quality of the CATHENA predictions under supercritical pressure conditions and the unique features which occur under these conditions. / Thesis / Master of Applied Science (MASc)

The Generation IV International Forum has established several goals for the next generation of nuclear energy systems, which are to be substantial improvements over contemporary designs. In Canada Generation IV research efforts have focused on developing the Pressure Tube type SuperCritical Water-cooled Reactor (PT-SCWR), an evolution of CANada Deuterium Uranium (CANDU) technology. An integral part of the PT-SCWR is the High Efficiency Re-entrant Channel (HERC), wherein coolant first travels downward through a centre flow tube and then upward around the fuel. The large density variation of supercritical fluids, combined with the negative Coolant Void Reactivity (CVR), make the concept similar to a Boiling Water Reactor (BWR). The objective of this study was thus to apply the state-of-the-art in BWR analysis to the PT-SCWR. Models were created using the DRAGON (neutron transport), DONJON (neutron diffusion/spatial kinetics), and CATHENA (channel thermalhydraulics) computer codes. A procedure for DONJON-CATHENA coupling was developed to enable simulation of coupled transients. The specifications of the HERC necessitated multiple coolant reactivity feedbacks be included in the model, in turn requiring extensions to the DONJON source code. The model created for this work is thus among the first to incorporate multiple coolant feedbacks in core-level coupled spatial kinetics and thermalhydraulics transient analysis, and is uniquely capable of simulating such transients in the PT-SCWR. This work found that while the total CVR was negative as required, the reactivity effect of coolant void solely around the fuel was positive. As a consequence additional heat delivered from fuel to coolant, which decreases the coolant density, has a positive reactivity effect making BWR-like coupled instabilities impossible. On the other hand, in some postulated transients, such as Loss-OfCoolant Accidents (LOCAs) or Loss-Of-Flow Accidents (LOFAs), this positive reactivity results in temporary power excursions. A fast-acting shutdown system is potentially necessary to limit damage to the fuel in such transients. / Dissertation / Doctor of Philosophy (PhD)

With an ever-increasing population both in Canada and globally, an improved quality of life will depend on having access to energy. The non-renewable, carbon-based, sources of energy that presently provide a major amount of the world's energy supply are depleting and therefore will be expensive in the future. Nuclear technology is a relatively new technology which can fulfill future energy needs but requires highly specialized skills and knowledge to continue to make it safer, cleaner, more reliable, and more affordable. Thus the nuclear industry puts lots of efforts to develop and improve the next generation of nuclear power plants. The Supercritical Water Reactors (SCWRs) are one of the Generation IV nuclear-reactor systems. The SCWRs, to a large extent, are very similar to light water reactors, but with a simpler design. The main advantage of SCWRs is their higher thermal efficiency. The Canadian SCWR has adopted an innovative fuel concept which is a mixture of plutonium and thorium oxides (Th, Pu) O2. The role of nuclear data in fuel development and reactor-physics analysis is quite significant. With the development of nuclear data files over the years, nuclear cross sections and other parameters are widely available, but their accuracy is still a concern. Also the accuracy of nuclear data is more reliable for uranium-based fuels than for thorium-based fuels. It is not known how the uncertainties in the nuclear data will impact the fuel depletion in a SCWR. Thus a sensitivity analysis tool has been developed to evaluate the impact of uncertainties in the neutron cross-sections of the actinides present in SCWR fuel. This document provides the details on the theory and methodology used to develop this tool (FICST). The objective of this work is to develop a code, not any specific calculation done with it. / Thesis / Master of Applied Science (MASc)

Canada has participated in the Generation IV International Forum (GIF) collaboration in the area of Super Critical Water-cooled Reactors (SCWR). Similar to the current CANDU technologies, in the Canadian SCWR design the low pressure heavy water moderator system is separated from the supercritical coolant system (25MPa). The High Efficiency Re-entrant Channel (HERC) design in the Canadian SCWR has multiple coolant regions (i.e. coolant in the downward center flow tube and coolant in the upward outer fuel region) and provides thermal isolation between the moderator and heat transport system fluid. Although the overall reactivity feedback in the channel is negative for equilibrium density decrease transients, a temporary positive reactivity may be induced during non-equilibrium conditions such as cold-leg Loss of Coolant Accidents (LOCA). The primary objective of this study is to investigate the fuel and coolant behaviors under postulated LOCA transients, in particular those caused by cold-leg breaks, and to demonstrate the effectiveness of several proposed safety systems in the Canadian SCWR. The one-dimensional thermalhydraulic system code RELAP5 has been used for the safety analysis in many LWRs. The latest version RELAP5/MOD4 has been improved to accommodate supercritical water and is used in this study. A RELAP5 model is constructed based on the most recent Canadian SCWR design. The 336 fuel channels are split into two representative groups each with a series of hydraulic and heat structure models. A benchmark study is conducted by comparing RELAP5 to CATHENA simulations and shows good agreement for both steady-state and transient predictions. The RELAP5 model is then used to predict the system response to several postulated LOCA transients. For a 100% (single-ended) cold-leg break located in between the feedwater pump and the inlet plenum, the system pressure immediately drops followed closely by a flow reversal with rapid discharge from the break. A brief power pulse (178%FP) is observed under this non-equilibrium depressurization scenario. The transient simulations show the potential for two sheath temperature maxima, one early in the transient as a result of the power pulse and the subsequent flow-power mismatch, and another later peak resulting from the fuel heat-up under near stagnant channel flow conditions (such as in the failure of the Emergency Core Cooling Systems) as the heat transfer regime changes to radiation dominated. The Automatic Depressurization System (ADS) located on the hot-leg side mitigates the later fuel heat-up by introducing forward channel flows. This effect is enhanced by additional coolant supplied from Low Pressure Coolant Injection (LPCI) which is part of the Emergency Core Cooling System (ECCS). Under the 100% break LOCA/LOECC transient, the core inventory is depleted rapidly after the break and thermal radiation becomes the dominant heat removal mechanism. The highest MCST, 1331 K, is achieved approximately 136s after the break and meets the safety criterion (1533 K). Beyond this time the sheath temperatures gradually decrease either by the continuous LPCI from the reactor sumps, gravity driven core cooling, or in the event of a failure of those systems by the Passive Moderator Cooling System (PMCS). LOCAs initiated by break sizes varying from 5% to 100% of the cold-leg cross-section area are simulated under loss of ECCS. In this specific design, break sizes less than 15% are defined as SBLOCAs and show an early pressure increase up to the Safety Relief Valve (SRV) setpoint. During SBLOCAs, the first MCST peak is more limiting than the large LOCA case because of insufficient fuel cooling caused by relatively low reverse flow. However, these lower reverse flows prolong the period of blowdown cooling and hence help to mitigate the secondary MCST peak. The worst LOCA case occurs in the 15% break case with a maximum cladding temperature of 1450 K. The results showed the most sensitive parameters are delays associated with SDS action, emissivity and ADS actuation parameters. / Thesis / Master of Applied Science (MASc)

The objectives of this thesis were the validation of G4STORK to use it for the investigation of the SCWR lattice cell. MCNP6 was chosen as the program that the methodology of G4STORK would be validated against. Over multiple steps, the methodology of G4STORK was matched to that of MCNP6 (described here, 3.4). After each step, the output of the two programs were compared, allowing us to pinpoint why and where discrepancies came about. At the end of this process, we were able to show that when G4STORK used the same assumptions as MCNP6, it produced similar results (shown here, 4.1.4). The results of G4STORK simulating the SCWR lattice cell, using its more accurate default methodology, was then compared to those of MCNP6 (shown here, 4.2.1). Large differences in the results were seen to occur, because of the inaccurate assumptions used by MCNP6, during transient cases. We concluded that despite the existence of minor discrepancies between the results of MCNP and G4STORK for some cases, G4STORK is still the theoretically more accurate method for simulating lattice cell cases such as these, due to MCNP’s use of the generational method. / Thesis / Master of Applied Science (MASc)

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7 – 20 kW/m2⋅K and 9.7 – 10 kW/m2⋅K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2, and MOX may reach melting point. / UOIT

A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT) - type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs. / UOIT

Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical Water-Cooled Reactor (SCWR) is to eliminate the annulus gap and use a ceramic insert to insulate the coolant from the moderator. While such a design may work, alternative fuel-channel design concepts are under development to explore the optimum efficiency of SCWRs. One such alternative approach is called the Re-Entrant fuel-channel. The Re-Entrant fuel-channel consists of three tubes, the inner tube (flow tube), pressure tube and an outer tube. The fuel bundles are placed in the inner tube. An annulus is formed between the flow and pressure tubes, through which the primary coolant flows. A ceramic insulator is placed between the pressure tube and the outer tube. The coolant flows through the annulus receiving heat from the inner tube from one end of the channel to another. At the far end, the flow will reverse direction and enter the inner tube, and hence the fuel-string. At the inlet, the temperature is 350°C for a high-pressure coolant (pressure of 25 MPa), which is just below the pseudocritical point. At the outlet, the temperature is about 625ºC at the same pressure (the pressure drop is small and can be neglected). The objective of this work was to design the Re-Entrant channel and to estimate the heat loss to the moderator for the proposed new fuel-channel design. A numerical model was developed and MATLAB was used to calculate the heat loss from the insulated Re-Entrant fuel-channel along with the temperature profiles and the heat transfer coefficients for a given set of flow, pressure, temperature and power boundary conditions. Thermophysical properties were obtained from NIST REFPROP software. With the results from the numerical model, the design of the Re-Entrant fuelchannel was optimized to improve its efficiency / UOIT

Due to the recent advancement in computer capability, numerical modelling starts to play an important role in making predictions and improving the understanding of physics in the studies of convective heat transfer to supercritical fluids. Many computational studies have been carried out in recent years to assess the ability of different turbulence models in reproducing the experimental data. The performance of these turbulence models varied significantly in predicting the heat transfer at supercritical pressures, especially for the phenomena of heat transfer deterioration (HTD). The results of these studies showed that the accuracy of different turbulence models was also dependent on the flow conditions. It is still necessary to test these turbulence models against newly available experimental data before the final conclusion can be drawn. In this work computational simulations on convective heat transfer of carbon dioxide (CO2) and water (H2O) at supercritical pressures flowing upward in vertical circular pipes have been carried out using the commercial code STAR-CCM+. Detailed comparisons are made between five turbulence models, including AKN low-Reynolds model by Abe et al. (AKN), Standard low-Reynolds k-ε model by Lien et al. (SLR), k-ω model by Wilcox (WI), SST k-ω model by Menter (SST), and the Reynolds Stress Transport (RST) model, against two independent experiments, i.e., water data by Watts (1980) and the recently published carbon dioxide data by Zahlan (2013). The performance of k-ε models with a two-layer approach, and that of k-ε models with wall-functions are also investigated. For the CO2 study, where wall temperatures in most cases are above the pseudo-critical temperature (Tpc), RST model is found both qualitatively and quantitatively better than other turbulence models in predicting the wall temperatures when HTD occurs. The RST model while superior, predicted HTD at higher heat fluxes as compared to experiments. The wall temperature trends predicted by SST and WI models are very similar to that predicted by RST, except that they start to predict HTD at even higher heat fluxes than RST, and the peak temperatures are overestimated significantly. Because RST and k-ω models (SST and WI) predict the HTD at higher heat fluxes as compared to experiments, often in literature they are overlooked. Rather CFD users should conduct sensitivity analyses on heat flux, and quite often as a result qualitatively excellent agreement can be observed in some of these models. The low-Reynolds turbulence models, i.e., SLR and AKN, tended to over-predict the wall temperature after the onset of first temperature peak, because the turbulence production predicted by these models failed to regenerate. The wall temperatures for these models did not show recovery after deterioration until the bulk temperature is close to Tpc, while experimentally recovery happened well upstream of this location. The k-ε models with two-layer approach, and the k-ε models with wall-functions both failed to predict the HTD in all cases. For the H2O study, where the wall temperatures in most cases are below the pseudo-critical temperature, the SLR model performed the best among all turbulence models in reproducing the experimental data. AKN model was also able to qualitatively predict the observed HTD, however, not as well as SLR. SST and RST models, on the other hand, under-predicted the buoyancy effect even at the lowest mass fluxes and hence did not adequately predict deterioration. In a few high-heat-flux cases with wall temperatures above Tpc, all the turbulence models show consistent response to that discussed in the CO2 study, i.e. RST model is quantitatively better than other turbulence models. Nevertheless, the wall temperature peaks predicted by RST model is very different from that observed experimentally, i.e. the measured peaks are much milder and more flattened than the predicted ones. All the turbulence models including RST overestimate the wall temperatures significantly when Tb<Tpc<Tw. The sensitivity studies of mesh parameters, user-defined fluid properties, turbulent Prandtl number, gravitational orientation, and various boundary conditions (e.g. heat flux, mass flux, pressure, and inlet temperature) have also been carried out, aiming to ensure the reliability of the obtained results, and to gain a deeper insight into the physics of heat transfer deterioration in supercritical fluids. Detailed mechanistic studies of HTD have been carried out for both the CO2 and H2O simulations using different turbulence models (RST, SST, and SLR) in various flow conditions. The radial distribution of fluid properties and turbulence at various axial locations provides direct evidence of the mechanisms involved near the locations of deterioration. The buoyancy effect is found to be responsible for the observed HTD in both experiments (i.e., when gravity forces are removed no deterioration is observed). The buoyancy force exerted on the near-wall low-density layer modifies the velocity profile (thus shear stress distribution) in a way that greatly reduces the near-wall turbulence production, resulting in the impairment of heat transfer. In the CO2 study where the wall temperature exceeds the Tpc in a very short distance from the inlet, the “entrance effect” is found to play a more important role in initially impairing the turbulence production. However, this effect is not observed in cases where wall temperature is below Tpc, which is attributed to the weaker density variation below Tpc. / Thesis / Master of Applied Science (MASc)

The research and development effort for the next generation of nuclear power stations is being coordinated by the Generation IV International Forum (GIF). The supercritical water reactor (SCWR) is one of the six reactor technologies currently being pursued by the GIF. The unique nature of supercritical water necessitates further examination of its heat transfer regimes. The GIF SCWR blind computational benchmark exercise is focused on furthering the understanding of the heat transfer to supercritical water as well as its prediction. A methodology for computational fluid dynamics (CFD) simulations using STAR-CCM+ 9.02.005 has been developed for submission to the GIF SCWR computational benchmark exercise. The experiments of the GIF SCWR computational benchmark exercise were those conducted by the Japan Atomic Energy Agency (JAEA). They are of supercritical water flowing upward in a 7 rod bundle. Of the three experimental cases there are (i) an isothermal case, (ii) a low enthalpy, low heat flux case and (iii) a high enthalpy, high heat flux case. A separate effects study has been undertaken and the SST turbulence model has been chosen to model each of the three experiments. A near wall treatment that ensures a y+<0.09 has been used for both of the heated cases and a near wall treatment that ensures a y+<0.53 has been used for the isothermal case. This computational approach was determined to be the optimal choice which balances solution accuracy with computation time. Final simulation results are presented in advance of the release of the experimental results in June 2014. / Thesis / Master of Applied Science (MASc)