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Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving VHTR Efficiency and Testing Material Compatibility 1. INTRODUCTION The U.S. and other countries address major challenges related to energy security and the environmental impacts of fossil fuels. Solutions to these issues include carbon-free electricity generation and hydrogen production for fuel cell car, fertilizer synthesis, petroleum refining, and other applications. The Very High Temperature Gas Reactor (HTGR) [INEEL, 2004] has been recognized as a promising technology for high efficiency electricity generation and high temperature process heat applications. Therefore, the U.S. needs to make the HTGR intrinsically safe and proliferation-resistant. The U.S. and the world, however, must still overcome certain technical issues and the cost barrier before it can be built in the U.S. The establishment of a nuclear power cost goal of 3.3 cents/kWh is desirable in order to compete with fossil combined-cycle, gas turbine power generation. This goal requires approximately a 30% reduction in power cost for state-of-the-art nuclear plants. It has been demonstrated that this large cost differential can be overcome only by technology improvements that lead to a combination of better efficiency and more compatible reactor materials. 1.1 Objective The objectives of this research are (1) to develop a supercritical carbon dioxide Brayton cycle in the secondary power conversion side that can be applied to some Generation-IV reactors such as the HTGR and supercritical water reactor, (2) to improve the plant net efficiency by using the carbon dioxide Brayton cycle, and (3) to test material compatibility at high temperatures and pressures. The reduced volumetric flow rate of carbon dioxide due to higher density compared to helium will reduce compression work, which eventually increase turbine work enhancing the plant net efficiency. 1.2 Background The VHTR is a graphite-moderated helium-cooled reactor that uses a direct or an indirect gas cycle to convert the heat generated by nuclear fission in the reactor and transferred to the coolant gas into electrical energy by means of a helium turbo-generator. The VHTR technology has been developed since the 1950s and has introduced the high-temperature gas-cooled reactor (HTGR) [IAEA, 1984]. HTGRs work on the principle of passing a cooling gas through the core, then running the heated gas directly to a steam generator or turbine. These reactors have been built in England and have been largely researched in Germany. The Arbeitsgemeinschaft Versuchsreactok (AVR), 15-MWe test reactor located in Forschungszentrum Juelich, was constructed starting in 1961. First criticality was achieved in 1966. The AVR has operated for 21 years. In 1974, the reactor outlet temperature was raised to 9500°C, which was needed to test very-high-temperature nuclear process heat applications. The most recent VHTR built is the Chinese HTR-10 (10 MW), which achieved first criticality December 2000 [Schleifenheimer, 1970, Gilli et al., 1975, Zierman, 1984]. HTR-10 [Hu et al., 2004] was designed to be operated up to 9500°C for investigating diverse power generation systems (e.g., gas turbine) and nuclear process heat applications [Generation IV Roadmap, 2001]. Recently, Eskom, a power company based in South Africa, submitted a nuclear installation license application to the National Nuclear Regulator (NNR). It proposes to locate the installation on Eskom property within the owner-controlled boundary of Koeberg Nuclear Power Station located in the Western Cape. In the United States, the Department of Energy will soon select one site out 1PDF Image | Development Of A Supercritical Carbon Dioxide Brayton Cycle
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