Development of a Computer Code for Thermo Hydraulics Analysis of Prismatic High Temperature Gas Cooled Reactors

Naderi, Mohammad Hossein | 2010

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  1. Type of Document: M.Sc. Thesis
  2. Language: Farsi
  3. Document No: 40143 (46)
  4. University: Sharif University of Technology
  5. Department: Energy Engineering
  6. Advisor(s): Ghofrani, Mohammad Bagher; Jafari, Jalil
  7. Abstract:
  8. A prismatic high temperature gas-cooled reactor (HTGR), which is a graphite moderated, helium-cooled reactor, is a promising candidate for next generation nuclear power plant in that it enables applications, such as hydrogen production or process heat for petrochemical by supplying heat with core outlet temperatures as high as 1000°C. A Thermal Hydraulic Analysis Code (THAC) for gas-cooled reactors has been developed. THAC implicitly solves heat transfer equation of fuel, graphite block and helium. Three types of fuel pins were considered; solid fuel pin, fuel pins with inside holes and annular fuels with coolant flow from its inside and outside surfaces. THAC predicts axial and radial temperature distributions across the reactor core by Orthogonal Collocation Method (OCM) and Finite Difference Method (FDM). Validation was carried out through comparison with experimental and analytical result. The predication agreed well with the experimental data. A computation module for annular fuel rods has been coupled to the code for comparative analyses of an annular fuel-based block element. At normal operation, the annular fuel shows lower peak temperature than the solid fuel for the same power in Japan’s high temperature engineering test reactor (HTTR), even though the pressure drop is higher in the annular fuel
  9. Keywords:
  10. Orthogonal Collocation ; Finite Difference Method ; Thermal Hydraulic Analysis Code (THAC) ; High Themperature Gas-Cooled Reactor (HTGR) ; Thermal Hydraulic Parameters

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