Thermal hydraulics analysis of supercritical water reactor core design Jianguo Zhong, Michael Podowski
Abstract: This paper started with a literature survey of development the supercritical flow reactor core design, various coolant flow path designs were discussed: conventional one-pass flow arrangement, two-pass and the three-pass coolant flow geometry. A simple 1-D model was developed for the one-pass and two-pass flow design to find the maximum flow bulk temperature and case temperature in the fuel assembly. In the implementation of the model, an accurate prediction of heat transfer coefficient was vital, which was also much complicated for the supercritical flow since the flow property changed substantially at the pseudo-critical region. Study showed that classical Dittus-Bolter correlation disagreed with the experimental results significantly. The in-house CFD code, NPHASE-CMFD, was implemented on 1-D supercritical water reactor to simulate the bulk flow and case temperature distribution in the supercritical water reactor, which provided useful guidance of supercritical water reactor design from engineering purpose. 1. Introduction
In the development of next generation nuclear power plant, higher efficiency and more compact size of reactor core is important. Among the six next generation nuclear power plant, supercritical water-cooled reactor (SCWR) are promising because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs ) and considerable plant component simplification with the direct cycle. In the SCWR, since the coolant in the core is operated above the critical pressure, no phase change process exists in the core in the heat transfer process, eliminating the boiling related issues such as critical heat flux and local dry out problem. Also, in the SCWR, there is no need for recirculation and jet pumps, a pressurizer, steam generators, and dryers. Lots of SCWR reactor core design concepts were developed in the past decades. There categories of these conceptual design will be introduced in the following section regarding the coolant flow arrangement in the reactor core, including one-pass flow, two-pass flow and three-pass flow arrangement.
One-Pass Flow Arrangement
In the conventional BWR reactor design, the coolant flow from the inlet nozzle is guided into lower core bottom from the down corner and flow upwards though the fuel assembly (Fig. 1a) The upward flow gets heated in the fuel assembly and leaves the reactor pressure vessel from the outlet nozzle. The advantage of the one-pass flow arrangement is its simple geometry. However, this one-pass flow arrangement would cause significant temperature non-uniformity of coolant flow among various fuel assemblies in the nuclear power plant due to non-uniform heat generation rate among fuel assemblies in radial and vertical direction. The presence of local hot spot could serious damage the structure of the reactor core. Thus, the in design of the reactor core, the core outlet temperature has to be decreased according to the local hot spot, thus, a decrease in nuclear power plant efficiency. Two-Pass Flow Arrangement
In order to obtain a more uniform outlet temperature and decrease the local hot spot temperature, Dobashi et al.  proposed a core conceptual design of two-pass coolant flow arrangement with hexagonal fuel assemblies and an extra moderator in the form of descending water rods named SCLWR-H. As shown in Fig 1.b, the supercritical coolant flow was separated into the upper core and lower core, then the coolant in the upper core flows downward in the water rod into the lower core and mixed with the lower core coolant before flowing upward through the fuel assemblies. During the downward flow process in the water reactor, the coolant in the water rod exchanges heat with the upward flow. The hexagonal fuel assembly design was later improved by MacDonald et al.  with square assemblies...
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