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A sodium-cooled fast reactor has been designed to attain a high burn-up core in commercialized fast reactor cycle systems. The sodium-cooled fast reactor adopts a wire spacer between fuel pins. The wire spacer performs functions of securing the coolant channel and mixing between subchannels. In high burn-up fuel subassemblies, the fuel pin deformation due to swelling and thermal bowing may decrease the local flow velocity in the subassembly and influence the heat removal capability. Therefore, understanding the flow field in a wire-wrapped pin bundle is important. This study performed particle image velocimetry (PIV) measurements using a wire-wrapped three-pin bundle water model to grasp the flow field in the subchannel under conditions, including the laminar to turbulent regions. The PIV results confirmed that the normalized flow velocity near the wrapping wire in the low Re number condition was relatively decreased compared to that in the high Re number condition. Meanwhile, in the region away from the wrapping wire, the maximum flow velocity was increased by decreasing the Re number. Accordingly, the PIV measurements using the three-pin bundle geometry without the wrapping wire were also conducted to understand the effect of the wrapping wires on the flow field in the subchannel. The results confirmed that the mixing due to the wrapping wire occurred, even in the laminar condition. These experimental results are useful not only for understanding the pin bundle thermal hydraulics, but also for the code validation.

Pronghorn-SC is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE). Initially designed to simulate flows in water-cooled, square lattice, subchannel assemblies, Pronghorn-SC has been expanded to simulate liquid-metal-cooled flows in triangular lattices, hexagonal subchannel assemblies. For this purpose, the algorithm of Pronghorn-SC was adapted to solve the subchannel equations as they are applicable to a hexagonal wire-wrapped sodium-cooled fast reactor. Cheng–Todreas models for pressure drop and cross-flow models were adopted and a coolant heat conduction term was added. To solve these equations, an improved implicit algorithm was developed robust enough to deal with the numerical issues, associated with low flow and recirculation phenomena. To confirm the prediction capability of Pronghorn-SC, calculations and comparisons with available experimental data of 19- and 37-pin assemblies were performed, as well as other subchannel codes. Finally, a flow blockage modeling feature was added. This capability was validated for both water-cooled square sub-assemblies and sodium-cooled hexagonal sub-assemblies, using experimental data of partially and fully blocked cases.

This paper investigates the admission control problem on the satellite multi-beam networks with non-orthogonal multiple access (NOMA). The goal is to maximize the number of supported users on the premise of ensuring the quality of service (QoS) by optimizing the subchannel and power allocation. We provide the system model and then formulate the admission control problem as a mixed integer non-convex optimization problem. The non-convexity and existence of integer variable make the optimal solution difficult to get. Therefore, we propose a joint subchannel matching and power allocation algorithm to obtain the suboptimal solution so as to reduce the computation complexity. The proposed algorithm can be used for both NOMA and orthogonal frequency division multiplexing access (OFDMA). Specifically, the subchannel matching problem is solved by a two-stage matching process where users are accessed to subchannel dynamically. The power allocation problem is modeled as a super-modular game where the existence and uniqueness of Nash equilibrium (NE) are analyzed. Moreover, an iterative power allocation algorithm is proposed based on the NE searching method. Finally, simulation results are provided for demonstrating the effectiveness and feasibility of the proposed algorithm.

The marine environment may change the force on the fluid and inevitably influence bubble behavior and the two-phase flow in the reactor core, which are vital to the safety margin of a nuclear reactor. To explore the effect of the marine motion on the flow and heat transfer features of subcooled flow boiling in the reactor core, the volume of fluid (VOF) method is employed to reveal the interaction between the interface structure and two-phase flow in the subchannel under rolling motion. The variations of several physical parameters are obtained, including the transverse flow, the vapor volume fraction, the vapor adhesion ratio, and the phase distribution of boiling two-phase flow with time. Sensitivity analyses of the amplitude and the period of the rolling motion were performed to demonstrate the mechanisms of the influence of the rolling motion. We found that the transverse flow in the subchannel was mainly affected by the Euler force under the rolling motion. In contrast to the two-phase flow in the static state, the vapor volume fraction and vapor adhesion ratio show different characteristics under rolling motion. Additionally, the onset of significant void (OSV) point changes periodically under rolling motion.

Currently, safety analyses mostly rely on codes which solve both the neutronics and the thermal-hydraulics with assembly-wise nodes resolution as multiphysics heterogeneous transport solvers are still too time and memory expensive. The pin-by-pin homogenized codes can be seen as a bridge between the heterogeneous codes and the traditional nodal assembly-wise calculations. In this work, the pin-by-pin simplified transport solver Tortin has been coupled with a sub-channel code COBRA-TF. The verification of the 3D solver of Tortin is presented at first, showing very good agreement in terms of axial and radial power profile with the Monte Carlo code SERPENT for a small minicore and with the state-of-the-art nodal code SIMULATE5 for a quarter core without feedback. Then the results of Tortin+COBRA-TF are compared with SIMULATE5 for one assembly problem with feedback. The axial profiles of power and moderator temperature show good agreement, while the fuel temperature differ by up to 40 K. This is caused mainly by different gap and fuel conductance parameters used in COBRA-TF and in SIMULATE5.

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. 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. 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. Master of Applied Science (MASc)

Turbulent mixing rates of gas and liquid phases between adjacent subchannels have been measured for various air-water two-phase annular flows in a multiple channel consisting of two identical circular subchannels. In order to study effect of liquid entrainment in the gas core on the turbulent mixing rates, experiments were conducted for two types of liquid injection method, i. e., a small bore nozzle placed in the subchannel center and a porous wall. The result showed that the effect of liquid entrainment on the turbulent mixing rates of both phases is negligibly small.