Thermal Hydraulic and Neutronic Analysis of a Reentrant Fuel Channel Design for Pressure Channel Supercritical Water Cooled Reactors

Abstract

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts, one of which is supercritical watercooled reactor (SCWR). Among the Generation IV nuclearreactor concepts, only SCWRs use water as a coolant. The SCWR concept is considered to be an evolution of watercooled reactors (pressurized water reactors (PWRs), boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), and lightwater, graphitemoderated reactors (LGRs)), which comprise 96% of the current fleet of operating nuclear power reactors and are categorized under Generation II, III, and III+ nuclear reactors. The latter watercooled reactors have thermal efficiencies of 30–36%, whereas the evolutionary SCWR will have a thermal efficiency of approximately 45–50%. In terms of a pressure boundary, SCWRs are classified into two categories, namely, pressurevessel (PV) SCWRs and pressurechannel (PCh) SCWRs. A generic pressurechannel SCWR, which is the focus of this paper, operates at a pressure of 25آ MPa with inlet and outlet coolant temperatures of 350آ°C and 625آ°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuelchannel design. Second, a nuclear fuel and fuel cycle should be selected. Several fuelchannel designs have been proposed for SCWRs. These fuelchannel designs can be classified into two categories: directflow and reentrant channel concepts. The objective of this paper is to study thermalhydraulic and neutronic aspects of a reentrant fuelchannel design. With this objective, a thermalhydraulic code has been developed in MATLAB, which calculates fuelcenterlinetemperature, sheathtemperature, coolanttemperature, and heattransfercoefficient profiles. A lattice code and diffusion code were used to determine a power distribution inside the core. Then, heat flux in a channel with the maximum thermal power was used as an input into the thermalhydraulic code. This paper presents a fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature exceeds the design temperature limits of 1850آ°C for fuel.