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The Canadian supercritical water-cooled nuclear reactor (SCWR) is a 2540 MWth channel-type SCWR concept that employs 336 fuel channels in the reactor core. Each fuel channel includes a pressure tube that is submerged in a heavy water moderator and contains a removable fuel assembly. The fuel assembly is designed so that all in-core components exposed to high radiation fields (other than the pressure tube) are part of the fuel assembly, which is removed from the reactor core as part of the assembly after three operating cycles. This design feature significantly reduces the likelihood of component failures due to radiation damage. To achieve high (>45%) power conversion efficiency, the Canadian SCWR operates at a supercritical water pressure (25 MPa) and high temperatures (350 °C at the inlet, 625 °C at the outlet). These conditions lead to fuel cladding temperatures close to 800 °C. Because of the reduced material strength at this temperature and higher fission gas production of the fuel, collapsible fuel cladding is selected over internally pressurized cladding. To increase heat transfer and to reduce cladding temperatures, turbulence-inducing wire-wraps are employed on fuel elements. Numerical models have been developed to analyze the thermal-structural behavior of Canadian SCWR fuel at normal and accident conditions. It was found that axial ridging, a possible failure mechanism with collapsed fuel cladding, can be avoided if the cladding thickness is larger than 0.4 mm. Detailed numerical analysis showed that the maximum fuel cladding temperature for the worst-case accident scenario is below the melting point by a small margin. This result was obtained with conservative assumptions, suggesting that the actual margin is greater. Hence, one of the design goals, the exclusion of the possibility of melting of the fuel, which is called the “no-core-melt” concept, seems attainable. However, this needs to be demonstrated more rigorously by removing the conservative assumptions in the analysis and performing supporting experimental work. This paper presents a description of the Canadian SCWR fuel assembly concept, its unique features, the rationale used in the concept development and the results of various numerical analyses demonstrating the performance and characteristics of the Canadian SCWR fuel channel.

A method to monitor the mechanical behavior and identify crack location and growth in a concrete structure element using a distributed fiber optic sensor (FOS) system is demonstrated experimentally by testing concrete specimens in four-point bending. The sensor system consisted of an optical frequency domain reflectometry (OFDR) interrogator unit paired with an all-grating sensing fiber that was bonded to the surface of the concrete test specimen. Strain measurements with high spatial resolution of <10 mm were obtained at various points along a single fiber cable. Large strain values at the crack locations indicated strain concentrations that could be used to assess the crack growth. The distributed sensing system demonstrated the capability to detect localized, early stage cracks, with crack width smaller than 0.1 mm, well before they become observable by visual inspection.

The Canadian Supercritical Water-Cooled Reactor (SCWR) is a 1200 MW(e) channel-type nuclear reactor. The reactor core includes 336 vertical pressurized fuel channels immersed in a low-pressure heavy water moderator and calandria vessel containment. The supercritical water (SCW) coolant flows into the fuel channels through a common inlet plenum and exits through a common outlet header. One of the main features of the Canadian SCWR concept is the high-pressure (25 MPa) and high-temperature (350°C at the inlet, 625°C at the outlet) operating conditions that result in an estimated thermal efficiency of 48%. This is significantly higher than the thermal efficiency of the present light water reactors, which is about 33%. This paper presents a description of the Canadian SCWR core design concept; various numerical analyses performed to understand the temperature, flow, and stress distributions of various core components; and how the analyses results provided input for improved concept development.