One of the challenges of small modular reactors (SMRs) in comparison with large reactors is the greater difficulty in achieving high burnups in smaller cores. With greater neutron leakage through the periphery, a key factor is the neutron economy of the fuel cladding. However, all large supercritical water-cooled reactor (SCWR) concepts have employed neutron-absorbing stainless steels and nickel-based alloys in order to meet all the requirements in terms of corrosion and thermalhydraulics. In order to achieve higher burnups and extend the time between refueling in a SCW-SMR, the use of chromium-coated zirconium alloy as a potential fuel cladding candidate has been explored. Chromium coatings up to a few micrometers thick have shown improved oxidation resistance of zirconium-based claddings under operating conditions relevant to SCWR concepts. In this study, Zr-2.5Nb alloy (UNS R60904) from pressure tube samples was coated using a physical vapor-deposition (PVD) method. Oxidation tests were performed on coated samples at 500 °C and approximately 25 MPa in a refreshed autoclave. The effects of the oxide on heat transfer and hydraulic resistance are also discussed in this study. Last, but not least, this study evaluates the coating cost of the fuel cladding with chromium in a vacuum plasma spray process.

The Canadian Nuclear Laboratories (CNL) is developing the technologies to enable the use of thorium-based fuels in pressure tube–heavy water reactors (PT-HWRs). One of the key stages in developing the thorium-based fuels for PT-HWRs is the reactor core configuration. Currently at CNL there are 20 core configurations under investigation, which involve several types of thorium-based fuels that could be implemented in a 700-MW(electric)-class PT-HWR. Among these core configurations, four fuel bundle concepts are being considered: (1) the reference (or nominal) 37-element bundle; (2) a 37-element modified bundle, with the center element using a different fuel material; (3) a 35-element bundle; and (4) an 18-element internally cooled annular fuel bundle. This study presents the steady-state subchannel thermal-hydraulic assessment of the 20 core configurations under investigation. The hottest channel approach is used in this study, as it represents the upper limit of a feasible design. The axial and element power distributions used in the analysis correspond to those of the discharge burnup. Three mass flows are considered in this study: 13.5, 21, and 24 kg/s. Five parameters are used to evaluate the fuel channel/bundle performance, namely, minimum critical heat flux ratio, channel pressure drop, enthalpy distribution, void fraction, and core power.

Canada has participated in the Generation IV (Gen-IV) International Forum with a main focus on the pressure-tube type supercritical water-cooled reactor (SCWR) concept. The Canadian SCWR concept is a heavy water moderated and light water cooled reactor. The research and development framework for development of the Canadian SCWR fuel-assembly concept includes experiments, analysis, subchannel code development, and applications, as well as coupled neutronic and thermalhydraulic analyses. The evolution of the Canadian SCWR fuel-assembly concept is summarized in this paper to demonstrate the enhanced performance characteristics and to show that the peak fuel cladding temperature is within allowed limits. The peak cladding temperature is a limiting thermalhydraulic parameter for the Canadian SCWR fuel assembly and is calculated using the ASSERT-PV subchannel code under steady state normal operating conditions with axial and radial power profiles corresponding to the beginning and the end of the fuel cycle. The... Le Canada a participé au Forum international Génération IV (GIF IV), qui portait surtout sur le concept de réacteur refroidi à l’eau supercritique (RESC) de type tube de force. Le concept de RESC canadien est un réacteur modéré à eau lourde et refroidi à eau légère. Le cadre de recherche et développement du concept d’assemblage de combustible du RESC canadien prévoit notamment des expériences, une analyse, l’élaboration du code de sous-canaux et des applications, ainsi que le couplage d’analyses thermohydrauliques et neutroniques. L’évolution du concept d’assemblage de combustible du RESC canadien est résumée dans le présent article afin de mettre en évidence les caractéristiques améliorées de rendement et de montrer que la température de pointe de la gaine de combustible se situe dans les limites permises. La température de pointe de la gaine est un paramètre thermohydraulique restrictif d’assemblage de combustible du RESC canadien et est calculée à l’aide du code de sous canaux ASSERT-PV dans des condit...

Canada is participating in the Generation IV International Forum with the main focus on the pressure-tube-type supercritical water-cooled reactor (SCWR) concept. The Canadian SCWR concept is a heavy-water moderated and light-water-cooled reactor. The R&D framework for the development of the Canadian SCWR fuel-assembly concept includes experiments and analyses, including subchannel code development and applications. This paper focuses on the modeling of the hydraulic resistance under supercritical conditions, with or without the wire-wrap spacers. More specifically, it presents an assessment of three friction factor correlations developed for supercritical conditions. A literature survey of wire-wrap hydraulic models is presented. The assessment of the supercritical friction factor correlations and wire-wrap hydraulic models is carried out using three pressure drop experimental datasets, using the subchannel code ASSERT-PV V3.2m2.

The thermal-hydraulics program in support of the development of the Canadian supercritical water-cooled reactor (SCWR) concept has undergone several phases. It focused on key parameters such as heat transfer, critical flow, and stability of fluids at supercritical pressures. Heat-transfer experiments were performed with tubes, annuli, and bundles in water, carbon dioxide (CO₂), or refrigerant flows. Data from these experiments have led to enhancement in understanding of the phenomena, improved prediction methods, and verified analytical tools. In addition, these experiments facilitated the investigation of separate effects on heat transfer (such as geometry, diameter, spacing device, and transient). Chocking flow characteristics were studied experimentally with sharp-edged nozzles of two different sizes of opening. Experimental data have been applied in improving the critical-flow correlation in support of accident analyses. A one-dimensional (1D) analytical model for instability phenomena has been developed and assessed against the latest experimental data for quantifying the prediction capability and applicability.