No description available.

Thesis (M. S.)--Mechanical Engineering, Georgia Institute of Technology, 2009. / Committee Chair: van Rooijen, Wilfred; Committee Member: Hertel, Nolan; Committee Member: Stacey, Weston

A series of accident simulations were performed using INL's thermal hydraulics code RELAP5-3D to analyze steady-state and transient behavior of a sodium cooled fast reactor. The reactor chosen for this study was General Electric's S-PRISM, which is a 1,000 MWt pool-type sodium-cooled fast reactor, designed for either an Oxide or Metal fueled core. Once key core characteristics including power profiles, reactivity feedback coefficients and delayed neutron parameters were calculated, S-PRISM was redesigned for a Nitride fueled core to take advantage of the Nitride fuel's high thermal conductivity and melting temperature. Loss of flow, loss of heat sink, loss of power and inadvertent control rod withdrawal accidents were simulated for each core at beginning, middle and end of cycle to determine if one fuel type provides significant safety advantages over the others.

Various fuel cycles for a sodium-cooled, subcritical, fast reactor with a fusion neutron source for the transmutation of light water reactor spent fuel have been analyzed. All fuel cycles were 4-batch, and all but one were constrained by a total fuel residence time consistent with a 200 dpa clad and structure materials damage limit. The objective of this study was to achieve greater than 90% burn up of the transuranics from the spent fuel.

Made available in DSpace on 2014-10-09T12:26:10Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:10:22Z (GMT). No. of bitstreams: 1 00951.pdf: 4937591 bytes, checksum: 160731d29ec9edf1fc78d0034f24638b (MD5) / Thesis (Doctorate) / IPEN/T / Massachusetts Institute of Technology - Cambridge, Mass - MIT

Made available in DSpace on 2014-10-09T12:26:10Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:10:22Z (GMT). No. of bitstreams: 1 00951.pdf: 4937591 bytes, checksum: 160731d29ec9edf1fc78d0034f24638b (MD5) / Thesis (Doctorate) / IPEN/T / Massachusetts Institute of Technology - Cambridge, Mass - MIT

Thesis (M. S.)--Mechanical Engineering, Georgia Institute of Technology, 2008. / Committee Chair: Willem F.G. Van Rooijen; Committee Member: Ghiaasiaan, Seyed M; Committee Member: Weston M. Stacey

As the United States expands its quantity of nuclear reactors in the near future, the amount of spent nuclear fuel (SNF) will also increase. Closing the nuclear fuel cycle has become the next major technical challenge for the nuclear energy industry. By separating the transuranics (TRU) from the SNF discharged by Light Water Reactors, it is possible to fuel Advanced Burner Reactors to minimize the amount of SNF that must be stored in High Level Waste Repositories. One such ABR concept is the Subcritical Advanced Burner Reactor (SABR) being developed at the Georgia Institute of Technology. SABR is a subcritical, sodium-cooled fast reactor with a fusion neutron source capable of burning up to 25% of the TRU fuel over an 8.2 year residence time. In the SABR concept an annular core with a thickness of 0.6 m and an active height of 3.2 m surrounds the toroidal fusion neutron source. Neutron multiplication varies during the lifetime of the reactor from keff = 0.95 at the beginning of reactor life to 0.83 at the end of an equilibrium fuel cycle. Sixteen control rods worth 9$ are symmetrically positioned around the reactor. This thesis describes the dynamic safety analysis of the coupled neutron source, reactor core and reactor heat removal systems. A special purpose simulation model was written to predict steady-state conditions and accident scenarios in SABR by calculating the coupled evolution of the power output from the fusion and fission cores and the axial and radial temperature distributions of a fuel pin in the reactor. Reactivity Feedback was modeled for Doppler and sodium coolant voiding. SABR has a positive temperature reactivity feedback coefficient. A series of accident scenarios were simulated to determine how much time exists to implement corrective measures during an accident before damage to the reactor occurs.